TSTF-21-03, TSTF Response to NRC Questions on TSTF-576, Revision 0, Revise Safety/Relief Valve Requirements and Submittal of Revision 1

TSTF Response to NRC Questions on TSTF-576, Revision 0, Revise Safety/Relief Valve Requirements and Submittal of Revision 1

ML21174A109
Person / Time
Site:	Technical Specifications Task Force
Issue date:	06/23/2021
From:	Joyce R, Miksa J, Demetrius Murray, Sparkman W, Vaughan J Technical Specifications Task Force
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TSTF-21-03
Download: ML21174A109 (80)

Text

TECHNICAL SPECIFICATIONS TASK FORCE TSTF A JOINT OWNERS GROUP ACTIVITY June 23, 2021 TSTF-21-03 PROJ0753 Attn: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" and Submittal of Revision 1 On August 27, 2019, the TSTF provided for NRC comment draft traveler TSTF-576, "Revise Safety/Relief Valve Requirements." A presubmittal meeting was held on September 12, 2019 and on October 21, 2019 the TSTF provided a revised draft. An additional presubmittal meeting was held on December 2, 2019.

On December 13, 2019, the TSTF submitted for NRC review Revision 0 of TSTF-576 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19347A726). On May 11, 2020, the NRC provided a Request for Additional Information (RAI) regarding TSTF-576 (ADAMS Accession Number ML19351D783). On August 11, 2020, the TSTF provided a draft RAI response and revised traveler for NRC comment. A teleconference to discuss the NRC's comments was held on October 13, 2020, followed by a teleconference audit on December 7, 2020.

On February 1, 2021, a revised draft RAI response and draft traveler were provided to the NRC for comment. The NRC provided comments on April 12, 2021 and a teleconference to discuss the NRC comments was held on May 13, 2021.

The TSTF's response to the NRC RAI and the additional comments is attached.

The RAI responses resulted in changes to TSTF-576. TSTF-576, Revision 1, is enclosed. The traveler includes markups based on the completed but not yet published Revision 5 of the Standard Technical Specifications.

11921 Rockville Pike, Suite 100, Rockville, MD 20852 Phone: 301-984-4400, Fax: 301-984-7600 Administration by EXCEL Services Corporation

TSTF-21-03 June 23, 2021 Should you have any questions, please do not hesitate to contact us.

James P. Miksa (PWROG/CE) Ryan M. Joyce (BWROG)

Dwi Murray (PWROG/W) Jordan L. Vaughan (PWROG/B&W)

Wesley Sparkman (PWROG/AP1000)

Attachment Enclosure cc: Michelle Honcharik, Technical Specifications Branch Victor Cusumano, Technical Specifications Branch

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" The NRC questions are repeated below in italics, followed by the TSTF response.

By letter dated December 13, 2019, the Technical Specifications Task Force (TSTF) submitted Revision 0 of Traveler TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements,"

(Agencywide Documents Access and Management System (ADAMS) Accession No. ML19347A726). TSTF-576, Revision 0, proposes to remove the safety relief valve (SRV) settings from the Standard Technical Specifications (STS) for boiling water reactor (BWR) plants. The setpoints are proposed to be moved to a licensee controlled in-service testing (IST) program and the limiting conditions for operation (LCOs) and surveillance requirements (SRs) for the related technical specification (TS) are proposed to be changed. The NRC staff has reviewed the submittal and determined that it does not contain adequate information to assure that regulations will be met if the submittal is adopted as written.

Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36 requires that TS be derived from the evaluations in the final safety analysis report (FSAR). It also requires that the TS include safety limits. The safety limit on reactor coolant system (RCS) pressure is a subject of the submittal. Section 50.36 of 10 CFR discusses limiting safety system settings for devices that have significant safety function and requires the limiting safety system settings are chosen such that automatic action will correct an abnormal condition prior to a safety limit being exceeded.

The regulation also requires that LCOs be established to define the lowest functional capability of equipment required for safe operation of the plant. With respect to the establishment of LCOs, the NRC staff cannot be assured that 10 CFR 50.36(b) will be met if the submittal is approved as written. These regulatory requirements must be considered individually and as whole to ensure that safe plant operation is assured. Additionally, the NRC staff concluded that the submittal does not provide adequate technical information to assure that Safety Limit 2.1.2 on RCS pressure will not be exceeded. Based on the submittal, the NRC staff could not conclude that the intent of 10 CFR 50.36 will be met if the change is approved as proposed. See the requests for additional information (RAIs) below for additional details. As discussed in the items below, the submittal does not provide adequate information to assure the staff that system testing, test frequencies, operability evaluations methods, and valve lift setpoints will be controlled adequately. The submittal also lacks technical bases for several statements included therein.

1. TSTF-576 states that future operability determinations will use assumptions similar to the overpressure analysis of record (AOR). The NRC staff understands that licensees will use licensing basis assumptions in their method for performing operability determinations.

Provide any differences from the design basis calculations that will be utilized in the operability evaluations. An obvious example of a change is the use of test results as inputs for SRV lift pressures instead of using the upper allowable existing TS setpoints. The NRC staff notes that the intent of FSAR Chapter 15 analyses is to use conservative assumptions to bound the results. Section 15.0 of the Standard Review Plan (NUREG-0800)1, directs the reviewer to verify that the applicant used parameters and initial conditions in the analyses that are suitably conservative, however, staff notes that as-found lift pressures for any given cycle are not conservative values.

1 U.S. Nuclear Regulatory Commission, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," NUREG-0800, Chapter 15, Section 15.0, "Introduction - Transient and Accident Analysis," Revision 3, March 2007 (ADAMS Accession No. ML070710376).

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TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" Response to RAI 1 The current S/RV LCO requires operability of each required S/RV. The current SR verifies each S/RV lifts within the tolerance around the specified setpoint. However, the specified safety function is based on the combined pressure relieving capacity of a group of the S/RVs and is not dependent on individual valve performance. The proposed change no longer specifies the number of S/RVs set at each lift setpoint and the as-found tolerance around the setpoint in the TS. Instead, the proposed change verifies the system level capability of the Overpressure Protection System (OPS) to prevent reactor steam dome pressure from exceeding Safety Limit 2.1.2.

Throughout these responses the term Reload Licensing Analysis (RLA) is used. This is the analysis performed prior to the start of a fuel cycle to confirm that the licensing basis assumptions (in this case, overpressure protection) will be met. Following their procedures, the licensee develops inputs to the RLA based on the analysis assumptions for the next cycle, such as trip setpoints, valve closure times, recirculation and feedwater pump performance, and turbine control valve response time. There are many input assumptions in the overpressure analysis in the RLA. The current TS specify the number, setpoint, and setpoint tolerance for the S/RVs.

However, there are other assumptions that are not specified in the current TS and are controlled by the licensee, such as the opening delay, stroke time, and closing pressure assumed for each S/RV, the S/RV capacity, and lengths and volumes of each steam line. The RLA is performed using the licensee's inputs and the NRC approved methods. For example, one General Electric methodology for evaluating overpressure events is TRACG, which was approved under NEDE-32906P-A, Revision 3, and subsequent revisions. Other vendors and licensees use other NRC-approved methodologies. The NRC-approved methodology for the licensee is documented in the UFSAR. These methodologies include conservatisms and margins to ensure the calculated values are bounding. Under the proposed change, compliance with the OPS LCO is confirmed using the NRC-approved methodology for the licensee.

The process to confirm the operability of the OPS under the proposed change is described in the following:

1) Cycle-specific overpressure analyses are performed as part of the Cycle N-1 RLA in accordance with the NRC-approved methodology as described in the UFSAR. The RLA demonstrates the capability of the OPS to protect Safety Limit 2.1.2 during Cycle N-1.

Currently the TS S/RV requirements are the inputs to the RLA overpressure analysis and the acceptance criterion for the analysis is Safety Limit 2.1.2. Under the proposed change, licensees will determine the S/RV overpressure analysis inputs and they can be revised to ensure they are conservative based on plant operating experience. The acceptance criterion continues to be the Safety Limit.

2) Following completion of Cycle N-1, periodic testing of S/RVs is performed as required by Appendix I of the American Society of Mechanical Engineers (ASME) Operations and Maintenance (OM) Code, "Class 1 Main Steam Pressure Relief Valves with Auxiliary Actuating Devices," Section I-3300, "Periodic Testing." 10 CFR 50.55a, "Codes and standards" requires licensees to follow the ASME OM Code.

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TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" a) Under OM Code Section I-1320, "Test Frequencies, Class 1 Pressure Relief Valves,"

all S/RVs must be tested at least every five years and a minimum of 20% of the S/RVs must be tested within any 24-month period2. There are two options: in-place testing and replacement with pretested S/RVs. If a subset of the S/RVs is replaced with pretested S/RVs, the S/RVs removed from service must be tested prior to restart. For each replaced S/RV that is tested for which the as-found opening pressure exceeds the established tolerance, two additional S/RVs must be tested. If any of the additional S/RVs open outside the established tolerance, the remaining S/RVs must be tested prior to startup. If the full complement of S/RVs is replaced with pretested S/RVs, the removed S/RVs must be tested within 12 months of removal.

Note that when S/RV setpoints are tested, only the pilot valve is tested. Generally, licensees that replace all S/RVs in lieu of testing replace only the pilot valves.

3) Following completion of the ASME-required testing, the results of the individual S/RV tests are evaluated against the Cycle N-1 RLA inputs and assumptions with the following potential outcomes:

a) If the as-found individual S/RV results are within the inputs and assumptions of the Cycle N-1 RLA (i.e., all tested S/RVs open within the assumed tolerance), no further action is required.

b) If the as-found individual S/RV performance is not within the inputs and assumptions of the Cycle N-1 RLA, the Cycle N-1 overpressure RLA is reevaluated using revised inputs that consider the as-found test results and the NRC-approved methodology for the licensee. The purpose of the reevaluation is to determine whether the Cycle N-1 SR was met. In addition, the reevaluation informs whether the current cycle SR must be examined as described below. The Cycle N-1 reevaluation may be performed by the licensee or a vendor. The reevaluation is performed using the measured S/RV lift settings for tested S/RVs. Any S/RV that was not tested will be considered to open within the upper limits of the established ASME Code testing acceptance criteria.

Any S/RV that was required to be tested but that could not be tested, or if the results cannot be determined, is assumed to be out-of-service. Trending and monitoring of S/RV performance is required under 10 CFR 50.65 (the Maintenance Rule).

The Cycle N-1 RLA reevaluation may result in one of the following outcomes:

i. If the Cycle N-1 RLA reevaluation demonstrates that the calculated overpressure is less than or equal to the Cycle N-1 RLA calculated peak overpressure for the limiting event (i.e., the RLA results were bounding), then the SR was met and the OPS was operable during Cycle N-1. It can be assumed that the current cycle RLA inputs and assumptions contain adequate conservatism to account for the as-found S/RV performance. No further action is required.

2 Which may be modified by Code Case OMN-17, Revision 1, "Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves," which is approved for unconditional use in Regulatory Guide 1.192, "Operation and Maintenance Code Case Acceptability, ASME OM Code."

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TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" ii. If the Cycle N-1 RLA reevaluation does not demonstrate that the calculated overpressure is less than or equal to the Cycle N-1 RLA calculated peak overpressure for the limiting event (i.e., the RLA results are not bounding), the issue will be entered into the Corrective Action Program (as required by 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action").

As part of the Corrective Action Program extent of condition consideration, the licensee will evaluate the effect of the Cycle N-1 as-found S/RV testing results on the basis for the Cycle N (i.e., the current operating cycle) RLA overpressure analysis inputs and assumptions. Changes to the inputs and assumptions will consider the cause of the as-found failures and the similarities or differences between the Cycle N-1 S/RVs and the Cycle N S/RVs. For example, if Cycle N-1 utilized two-stage S/RVs and three-stage S/RVs were installed for Cycle N, no changes to the inputs may be needed.

i) The reevaluation may determine that the Cycle N RLA is bounding (the calculated peak overpressure is less than or equal to the RLA calculated peak pressure). This demonstrates that the OPS is operable for Cycle N.

ii) The reevaluation may determine that the Cycle N RLA is not bounding (the calculated peak overpressure is greater than the RLA calculation) but that the OPS is operable (i.e., the Safety Limit would be protected in the limiting event). In this case, the OPS is operable for Cycle N. However, this determination will result in additional actions, such as updating of the Cycle N RLA, an evaluation of the effect of the higher calculated pressure on other analyses, and consideration of the 10 CFR 50.72 and 10 CFR 50.73 reportability requirements.

iii) The reevaluation may determine that the Cycle N RLA is not bounding and that the Safety Limit would not be protected in the limiting event. In that case, the OPS is inoperable, and a plant shutdown is required by the TS.

a. Discuss how the operability determination method will remain conservative and assure sufficient margin to the RCS pressure Safety Limit is maintained as required by Criterion 15, "Reactor Coolant System Design," of Appendix A "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50. Consider including a minimum margin between a licensees operability determination value and the Safety Limit in the revised TS Bases.

Response to RAI 1.a The RLA evaluates overpressure protection for the upcoming fuel cycle and verifies that the OPS will be operable and Safety Limit 2.1.2 will be protected. The RLA uses NRC-approved methods which include accepted conservatisms and assumptions. If a reevaluation of the current cycle RLA overpressure analysis (Cycle N) is required, the NRC-approved methods (including the approved conservatisms and assumptions) will be used. Therefore, additional margin does not need to be included.

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TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements"

b. Discuss whether the methodology will use the AOR assumptions (other than the use of tested lift pressures in place of assumed lift pressures).

i. Discuss whether as-tested lift pressures that are lower than the assumptions in the AOR will be used. For example, if a valve lifts below the setpoint, or below the assumed lift point used in the AOR, what value is used in the operability evaluation?

ii. Describe whether the methodology assumes any valves out of service (unable to lift).

Compare this to existing AOR assumptions.

iii. Provide a description of what will be done in the analyses for any valve that is not tested.

iv. Provide similar information for valves that are tested, but the lift setpoint cannot be determined.

v. Discuss the ability of the methods used to model valve flow characteristics and line losses for the use of multiple SRVs, specifically any changes that might be needed in order to account for each individual valves opening setpoint.

Response to RAI 1.b

i. Actual lift pressures, whether higher or lower than the RLA inputs and assumptions, will be used for tested S/RVs to determine whether the RLA is bounding.

ii. The assumptions related to out-of-service S/RVs vary based on licensee and by vendor methodology. Changes to the assumptions regarding S/RVs out of service would be considered changes to the facility and evaluated under 10 CFR 50.59.

iii. Any S/RV that was not tested will be considered to open within the upper limits of the established ASME Code testing acceptance criteria.

iv. Any S/RV that is required to be tested per the ASME OM Code but that cannot be tested or the results cannot be determined would be assumed to be out-of-service.

v. The GEH transient analysis methodology, both ODYN and TRACG, and the Framatome transient analysis methodology, both COTRANSA-2 and AURORA-B, incorporate S/RV modeling that can simulate individual valve lift / reseat setpoints, relief capacity, opening /

closing profiles, and inlet line losses with no other changes to the overall plant model. In addition, any number of valves out-of-service can be postulated as necessary.

c. Discuss whether single failure assumptions other than those used in the AOR should be evaluated. For example, is the failure of an SRV to open more or less conservative than a delayed scram due to failure of an anticipatory trip.

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TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" Response to RAI 1.c Single failure assumptions will continue to be consistent with the NRC-approved methodologies and are not affected by the proposed change. The NRC-approved methodologies are transient in nature and fully account for the timing of the event. The limiting analyzed overpressure event is the closure of all main steam isolation valves (MSIVs) followed by a reactor scram on high neutron flux, and the most limiting single failure is the failure of the direct scram associated with MSIV position. The assumption that an S/RV fails to open is not the assumed single failure in the analysis and is not a regulatory requirement. The typical analysis assumption that one S/RV fails to open is included for operational flexibility as is simplifies the evaluation should an S/RV fail to open within tolerance during as-found testing.

d. TSTF-576 states that future operation with the reinstalled SRVs is expected to meet all design and licensing basis requirements. Provide the basis for this statement and for using a methodology that incorporates only previous cycle test data in a forward-looking evaluation. Specifically, discuss why there would be reasonable assurance that the forward-looking evaluation will meet the design and licensing basis requirements. Include consideration of the degradation mechanism affecting the valve lift setpoints.

i. Provide a justification for the valve lift data used in the operability determinations.

How was it determined that the removed and tested valve lift pressures assure acceptable performance for the next cycle? If valves are not removed during each refueling outage how will it be determined that they will continue to operate acceptably?

ii. Describe how uncertainties in the data due to the degradation mechanism, test methods, and other factors, are considered in the use of the test data.

iii. Justify the statement that "Future operation with the reinstalled S/RVs is expected to meet all design and licensing basis requirements" when it is known that they frequently drift to setpoints outside the design and setpoints assumed in the safety analysis.

iv. TSTF-576 states "In all cases in which the SR was not met due to setpoint drift, the Licensee Event Reports (LER) concluded that the S/RVs as a group would have retained the capability to protect Safety Limit 2.1.2." This statement appears to be the basis for the statement that all SRVs are expected to meet all design and licensing basis requirements. Staff is under the impression that the LER findings are based on best-estimate calculations. Would the same findings have been made if conservative AOR calculations were performed?

Response to RAI 1.d The statement that "future operation with the reinstalled S/RVs is expected to meet all design and licensing basis requirements" was based on past industry operating experience. However, the TSTF does not have access to the methods or assumptions that licensees used to reach this conclusion and it is unclear whether the past operability analyses were best-estimate or used NRC-approved, licensing-basis methods. Licensees and their vendors are capable of performing 6

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" such analyses to support the conclusion, but absent detailed information on those past evaluations, the statement is removed.

Under the proposed change, any evaluation of OPS operability will be performed using NRC-approved methods for the licensee, including any conservatisms and assumptions. The acceptance criteria for the overpressure analysis is not affected by the proposed change.

Therefore, the licensing basis limits will continue to be protected.

i. The periodic testing of the S/RVs is performed in accordance with the ASME OM Code.

Individual S/RVs are tested per the ASME OM Code to determine the as-found lift setpoint.

This may be representative testing or testing of every valve. Following as-found testing, the S/RVs are refurbished, and as-left testing is performed in accordance with the ASME OM Code. The testing has been approved by the ASME, and endorsed by the NRC in the regulations, as being adequate and appropriate. Changes to the S/RV testing requirements are outside the scope of the proposed change.

ii. The periodic testing of the S/RVs is performed in accordance with the ASME OM Code.

Test methods are in accordance with the ASME OM Code and are outside the scope of the proposed change. Any required evaluation will be performed using NRC-approved methods and would consider uncertainties in the data in a manner consistent with licensee procedures.

iii. See response to RAI 1.d. The drift of individual S/RV setpoints to outside the assumed tolerance has not been shown to prevent the collective ability to protect the Safety Limit, which is the licensing and design basis of the TS requirement.

iv. See response to RAI 1.d. The TSTF cannot speculate on the results of analyses performed by multiple licensees and vendors if different assumptions were used.

e. Considering the importance of previous cycle data in establishing current operability, discuss the timing associated with performing any required past operability analyses based on the removed SRV test results. Provide a recommended time for completing the past operability determination, if it is required. Provide an example of a past operability evaluation that illustrates how an evaluation would typically be performed.

Response to RAI 1.e See the response to RAI 1. The time limits associated with the performance of S/RV testing are given in the ASME OM Code, as endorsed by the NRC in 10 CFR 50.55a. Unless all of the S/RVs are replaced with pretested valves, the testing must be performed before startup. If all of the S/RVs are replaced with pretested valves, the removed valves must be tested within 12 months. Changes to the timing of the ASME OM Code testing are outside the scope of the proposed change.

The timing of any required past operability evaluation is determined by the requirements of 10 CFR 50.72 and 50.73 and licensee's procedures. The proposed change does not affect the performance of past operability evaluations.

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TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" Evaluations of current operability of the OPS based on the results of the S/RV testing are performed under the Corrective Actions Program. 10 CFR 50 Appendix B, Criterion XVI, "Corrective Actions," establishes the requirements for prompt identification and correction of conditions adverse to quality.

There is no single example of past operability evaluations performed by licensees that would be typical. Licensees perform past operability evaluations following their procedures, frequently with support of a vendor, and the type of the evaluation will depend on the S/RV test results.

The response to RAI 1 provides a general description.

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TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements"

2. 10 CFR 50.36 (b) states "The technical specifications will be derived from the analyses and evaluation included in the safety analysis report" Currently, there is consistency between the TSs and analyses in the safety analysis report as the SR setpoint range is the same as that used in the AOR. However, with the new proposed TS, this will not be the case as the post-SR calculations will compare peak dome pressure with the safety limit rather than the setpoints used in the AOR. Therefore, the proposed TS changes do not appear to be consistent with the current analysis as described in licensees FSARs. Provide a discussion of how the proposed TS changes are consistent with existing plant AORs. How will plants AORs remain consistent with the methodology used to determine operability.

a. Describe how a licensee would proceed if the operability evaluation determines that the calculated peak RCS dome pressure does not exceed the Safety Limit, but the value in the plants AOR is exceeded.

b. Does the AOR need to be revised to account for system level performance instead of component level?

c. Does the AOR need to be revised to reflect a reduced margin to the TS Safety Limit?

d. Discuss how SRVs that are found to lift outside the assumptions in the AOR and/or the established as-found tolerance will be evaluated for an unanalyzed condition that significantly degrades plant safety.

Response to RAI 2 The purpose of the TS requirement is to ensure that the output of the RLA overpressure analysis verifies that the steam dome pressure will be prevented from exceeding Safety Limit 2.1.2 in the limiting event. The safety analysis acceptance criteria are based on whether the ASME OM Code overpressure limit, stated in Safety Limit 2.1.2, will be protected. The proposed TS aligns with that purpose. Therefore, the proposed change is more consistent with the safety analysis than the current TS, which specifies the inputs to the analysis and not the acceptance criteria.

Under the proposed change, determinations of whether the OPS is operable, and the SR is met will be performed using the NRC-approved methods approved for the licensee as documented in the UFSAR. These methods contain margins and conservatisms approved by the NRC. No changes to the methods are requested or required. Evaluation of as-found conditions will affect the inputs to the methods but will not affect the calculation or acceptance criteria.

Response to RAI 2.a As described in the response to RAI 1, if the as-found S/RV lift setpoints are not within with the assumptions in the previous cycle RLA, a reevaluation of the previous cycle RLA overpressure analysis must be performed to determine if the previous cycle analysis bounded actual plant performance. This reevaluation will consider the as-found S/RV lift setpoints. If the reevaluation does not bound the previous cycle analysis but the calculated peak pressure is less than the Safety Limit, the OPS was operable during the previous cycle. If the reevaluation does not bound the previous cycle analysis, the current cycle RLA inputs and assumptions must be examined to determine if the current cycle overpressure analysis must be reevaluated.

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TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" Response to RAI 2.b The RLA overpressure analysis evaluates system performance using NRC-approved methods.

Changes to the analysis inputs do not affect the use of the NRC-approved analysis method.

Therefore, the proposed TS change does not require a change to the analysis-of-record method.

Response to RAI 2.c The proposed change does not reduce the margin to the safety limit. The RLA is performed using the same NRC-approved methods, including any conservatisms or assumptions, and verifies that the overpressure safety limit is protected during the assumed accidents and transients.

Response to RAI 2.d See the response to RAI 2.a. The ability of the OPS to protect the overpressure Safety Limit is based on the system performance, not on the performance of individual S/RVs. If an S/RV does not lift within the assumed tolerance, the system performance is reevaluated to determine if the safety function can be performed. Only if the system performance fails to protect the Safety Limit would there be an effect on plant safety.

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TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements"

3. TSTF-576 states that the measured SRV lift pressures will be used in the analysis to determine if the SR is met. This implies that all SRVs will be tested, however, the traveler also states that periodic testing of SRVs will still be performed as required by Appendix I of the ASME OM Code. Discuss this discrepancy and describe what will be done to determine values for valve lift pressures to be used in the analysis.

Response to RAI 3 The traveler has been revised to clarify that the measured as-found S/RV lift settings will be used in the analysis for tested S/RVs and the upper limits of the ASME Code testing allowance will be used in the analysis for any S/RVs that were not tested.

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TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements"

4. Provide a justification that the change is adequate to assure that the SRVs open when manually actuated and that the downstream piping is unobstructed. Alternately justify that this action is not required or will be performed during post-maintenance testing.

a. The existing BWR/4 SR 3.4.3.2 requires that the valves open when manually actuated.

The BWR/6 SR 3.4.4.3 is identical. The existing STS Bases for these SRs state that the purpose of the SR is to verify that the valve is functioning properly and that there is no blockage in the discharge line. The proposed change to BWR/4 SR 3.4.3.2 adds a note that valve actuation may be excluded, and the wording of the SR is changed to require that only the valves acting in the relief mode actuate on an initiation signal.

The note in the proposed SR is contradictory to the SR requirement. The SR requires valve actuation, but the note says actuation may be excluded. This is confusing and should be changed to provide clear guidance. The proposed STS Bases state that the purpose is to verify that the mechanical portions of the auto relief valve mode operation. The SRV must open to perform these functions and opening is a mechanical action. The change to BWR/4 SR 3.4.3.2 and the deletion of BWR/6 SR 3.4.4.3 eliminate the requirements that assure that the valves will actually open and be able to pass flow. What is the basis for revising the TS to eliminate the requirement to actuate all SRVs? Provide a justification for eliminating the SRs that clearly require each valve be actuated and flow through the discharge piping be verified.

b. If these functions are being verified via post-maintenance testing, justify their removal as SRs and provide the actions that plants will take to verify that the valves will mechanically actuate and that the discharge lines are free of obstructions.

Response to RAI 4.a and 4.b Existing BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3 state:

---

NOTE------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify each [required] S/RV opens when manually actuated.

The Bases for the SRs state that a manual actuation is performed to verify that, mechanically, the valve is functioning properly, and no blockage exists in the valve discharge line. This SR is removed from the TS.

The STS SR is not representative of the plant-specific TS. Ten of the thirty BWR units have no equivalent SR. An additional thirteen of the thirty units have different requirements, such as verifying the actuator strokes (eleven units) or verification that the valve is capable of being opened (two units). Only seven of the thirty units contain an SR equivalent to the STS SR.

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TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" There is no safety analysis assumption that the safety mode S/RVs will open when manually actuated to limit overpressure. As a result, the ability to open manually is not required to demonstrate that the valve is functioning properly. This is supported by the existing SR Bases statement, "If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the S/RV is considered operable." TS SR 3.0.1 states that when an SR is not met, the associated LCO is not met (i.e., the system is inoperable).

However, the existing TS Bases are contrary to SR 3.0.1 and state that the LCO is met even if the manual actuation SR is not met. The manual actuation SR is removed to be consistent with the TS requirements.

BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3 are not needed to verify the S/RVs are mechanically functioning properly. The as-left test of the S/RVs verifies the mechanical functioning of the valve. Following installation in the plant, the actuator is tested with the actuator disengaged.

This test not only verifies operability of the actuator, but also verifies the pneumatic and electric connections. The actuators are then re-engaged to the valve with appropriate independent verification and sign-offs to verify the S/RV will function properly. Mechanically opening the S/RV after installation is avoided as it can result in seat leakage. The STS does not require post-maintenance or post-installation testing, as described in the SR 3.0.1 Bases as that is a maintenance activity carried out before declaring the component operable, and SRs are not required to be performed on inoperable equipment. Therefore, an SR is not needed for that purpose.

The SR is not needed to verify the downstream piping is unobstructed. Licensees have robust Foreign Material Exclusion (FME) programs to ensure systems are not contaminated during maintenance. Those programs are routinely inspected under NRC Inspection Manual Procedure 71111.20. Further, in licensing actions the staff agreed that an SR to verify that S/RV downstream piping is not obstructed is unnecessary because of licensee FME controls. The safety evaluation for license amendment 116 for the Hope Creek Generating Station, dated February 10, 1999, stated:

Another difference between the current TS-required stroking and the licensee's proposal is that, when performing the testing in-situ as required by the current TS, the testing verifies that the SRV discharge line is not blocked. However, the licensee stated that there is a Foreign Material Exclusion Program in place at the plant which minimizes the potential of debris blocking the discharge lines such that the possibility of blockage is extremely remote. The staff agrees that there is a very small possibility of blockage of an SRV discharge line as demonstrated by operational history and finds that the licensee has acceptably addressed this concern.

In addition to Hope Creek, there are nine additional BWR unit TS that do not require manual actuation of the safety mode S/RVs. There is no industry operational history of S/RV downstream piping being obstructed by foreign material and the staff's conclusion regarding the Hope Creek FME program is equally applicable to other plants adopting the proposed change.

In conclusion, the SR to perform a manual actuation of the safety mode S/RVs is not necessary to ensure the specified safety function can be performed.

13

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" Existing BWR/6 SR 3.4.4.2 states:

---

NOTE------------------------------

Valve actuation may be excluded.

Verify each [required] relief function S/RV actuates on an actual or simulated automatic initiation signal.

Testing of S/RV actuation in relief mode is unaffected by the proposed change. As discussed in the traveler, all four BWR/6 plants and two non-BWR/6 plants (Dresden 2 and 3 and Quad Cities 1 and 2) are permitted to credit some S/RVs in relief mode in addition to the S/RVs in safety mode to meet the ASME Code overpressure limit. In relief mode, actuators on the S/RVs are used to open the valves. The initiating instrumentation of the S/RVs in relief mode will continue to be tested by BWR/6 TS 3.3.6.5, "Relief and Low-Low Set (LLS) Instrumentation." The two BWR/4 plants that credit S/RVs in relief mode will retain similar requirements.

The existing SR Note states that S/RV actuation is not required when verifying that the S/RV will open on an actual or simulated automatic initiation signal. Most plants perform the SR by removing the actuator from the S/RV to avoid opening the S/RV. Opening of an S/RV after installation is avoided because it can lead to inadequate seating and seat leakage.

Revision 0 of TSTF-576 proposed to add an optional SR equivalent to BWR/6 SR 3.4.4.2 to the BWR/4 TS. Because the optional SR would only be applicable to two non-BWR/6 plants, the addition was removed from Revision 1 of the traveler.

14

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements"

5. The requested change focuses solely on the overpressure protection function of the SRVs.

Provide a discussion of how the change will ensure that trends in valve behavior will be evaluated to ensure that plant safety is adequately assured. For example, how will valve test results be trended and evaluated? Will that be done at an individual plant level only, or at a fleet or industry level? What, if any, visibility will the NRC have to that trending data.

a. Opening at a lower pressure has the potential to initiate a transient and challenge other safety systems. Provide a justification for removal of the lower setpoint from TS.

b. Removal of the valve tolerances from TS could ultimately result in less testing and maintenance applied to the valves which in turn might result in an increased tendency for the SRVs to drift to lower setpoints and leak. Seat leakage can contribute to damaging water hammer following a loss of offsite power during restart of the residual heat removal system as described in NRC Information Notice 87-10, Supplement 1, "Potential for Water Hammer during Restart of Residual Heat Removal," dated May 15, 1997. Describe how this issue was considered in development of the traveler and how it will be prevented in the future.

Response to RAI 5 S/RVs fall under 10 CFR 50.65 (the Maintenance Rule). Licensee Maintenance Rule programs require establishing performance criteria, monitoring and trending performance, determining the cause of failures, and taking corrective action. Those activities are available for NRC inspection.

In addition to these licensee-specific activities, the Boiling Water Reactor Owner's Group (BWROG) has been working to improve S/RV performance for many years and has trended the performance of problematic two-stage S/RVs. The BWROG plans to continue to monitor and improve S/RV performance across the BWR fleet. The BWROG routinely discusses these activities with the NRC.

The proposed change focuses on the overpressure protection function of the S/RVs because that is the purpose of the affected TS. The LCO Bases for BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4 state that the requirements of the LCO are applicable only to the capability of the S/RVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded. That purpose is not altered under the proposed change.

Response to RAI 5.a The proposed change does not justify licensee-specific alteration of the S/RV setpoints and tolerances and only moves the existing setpoints and tolerances to licensee control. Any subsequent change to the setpoints and tolerances must be evaluated under 10 CFR 50.59, which will consider the potential of the change to increase the probability of an accident or to create a new kind of accident.

Note that some BWR plants have NRC-approved S/RV tolerances of +3%/-5% and have not experienced inadvertent actuations.

15

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" Response to RAI 5.b The ability to perform the specified safety function by the current and proposed LCO is only dependent on the ability to relieve excess pressure and maintain reactor pressure below Safety Limit 2.1.2. The ASME O&M Code testing verifies that the S/RVs open within the lower and upper tolerance, but the safety function is only dependent on the S/RVs opening within the upper tolerance. Under the proposed change, the ASME O&M Code testing of the S/RVs will continue to be performed. Should an S/RV open below the lower tolerance, the test would be considered a failure and evaluated under the licensee's Corrective Action Program.

S/RV seat leakage is not governed by the existing or proposed BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4, and is not affected by the proposed change.

16

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements"

6. Current TS have LCOs for one or more SRVs out of service. The current STS Bases state that all SRVs must be operable due to the energy in the system. The STS Bases also state, with one or two SRVs inoperable the overall reliability of the SRV system is reduced. Thus, the LCO [sic] for one or two valves inoperable exists. The traveler does not address the effects of the inoperability of one or more SRVs on the overall reliability of the system. This is similar to having one train of a safety system inoperable. Considering that the valves are less reliable than originally assumed it seems that the consideration of reliability is more relevant than when the existing TS were developed. Provide the basis for deleting these LCOs [sic] from 3.4.3 (BWR/4) and 3.4.4 (BWR/6). Discuss how allowing one or more valves to remain inoperable while the overpressure protection system is operable maintains sufficient margin as required by Criterion 15.

Response to RAI 6 In the question, references to LCOs appear to be referring to Actions.

The current standard TS include an Action applicable when one or two required S/RVs are inoperable. The Action requires the inoperable S/RV(s) to be restored within 14 days, followed by a plant shutdown. The proposed TS does not include such an Action and S/RVs may be nonfunctional if the Overpressure Protection System is operable.

NUREG-1433 and NUREG-1434 contain an Action for one or two inoperable required safety mode S/RVs, but no BWR plant TS contain equivalent requirements. Three plants (Hatch, Hope Creek, and Monticello) permit operation with one S/RV inoperable, but the corresponding LCOs require more S/RVs to be operable than are credited in the overpressure analysis. As a result, for those plants the assumptions of the overpressure analysis can still be met when one or two safety mode S/RVs are inoperable. This is consistent with the STS Required Action A.1 Bases which states, "With the safety function of one [or two] [required] S/RV[s] inoperable, the remaining OPERABLE S/RVs are capable of providing the necessary overpressure protection. Because of additional design margin, the ASME Code limits for the RCPB can also be satisfied with two S/RVs inoperable. However, the overall reliability of the pressure relief system is reduced because additional failures in the remaining OPERABLE S/RVs could result in failure to adequately relieve pressure during a limiting event. For this reason, continued operation is permitted for a limited time only." (Note that Hatch was the lead plant for the development of NUREG-1433 and this is why the Required Action Bases reflect the Hatch requirements.)

The remaining BWR plants' current LCOs only require operability of the number of S/RVs assumed in the overpressure analysis. There are no additional S/RVs required to be operable.

Therefore, if one of the required S/RVs is inoperable, the assumptions of the overpressure analysis cannot be met and a plant shutdown is required. Similar to those plants' TS, the proposed Overpressure Protection System LCO does not require additional S/RVs beyond what is required to perform the overpressure protection function, Required Action A.1 is not applicable, and it is removed.

The existing BWR/6 TS 3.4.4 Actions are not consistent with the plant TS of the four BWR/6 plants. TS 3.4.4 for the BWR/6 plants (River Bend, Grand Gulf, Perry, and Clinton) contains a single action for one or more required safety mode or relief mode S/RVs inoperable, and it 17

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" requires being in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, which is consistent with the proposed change. Therefore, the BWR/6 STS is revised to be consistent with the plant-specific BWR/6 TS.

Under the proposed change, the TS LCO requires the OPS to be operable. The LCO and SRs do not place requirements on individual S/RVs. (The number of S/RVs credited in the overpressure analysis is described in the Bases.) Therefore, there is no condition in which the OPS could be inoperable (i.e., the LCO would not be met and the Actions apply) but the remaining S/RVs could be capable of providing the necessary overpressure protection to protect the Safety Limit as stated in the current Action Bases. The only appropriate Action when the proposed LCO is not met is to shut down the unit.

An inoperable required S/RV is not analogous to an inoperable train in a safety system. Safety systems have independent and redundant trains and each train is capable of performing the safety function. The S/RVs cannot individually perform the overpressure protection function but must function as a group. Not all S/RVs may be needed to protect the overpressure safety limit.

Many BWR plants have more S/RVs installed than are required by the current TS LCO. For these plants, the LCO can still be met (and no Action is applicable) for one or more nonfunctional S/RVs if the required S/RVs are operable. For example, if a plant had nine installed S/RVs and the LCO required seven S/RVs to be operable, three (not one) S/RVs would need to be incapable of performing their function before the STS LCO would not be met and the Actions would apply.

This relaxation is acceptable because the purpose of the LCO is to protect the overpressure Safety Limit and if the LCO can be met, it is unnecessary to limit continued plant operation due to the unavailability of one or two S/RVs. However, should it be determined that an S/RV is unavailable, an evaluation under the Corrective Action Program would be required to determine whether the Overpressure Protection System is operable. Availability of the S/RVs will also be assessed under 10 CFR 50.65 (the Maintenance Rule).

18

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements"

7. Page 4 of the model application, response to question 1, states that the ability of the SRVs to mitigate any accident previously evaluated is not significantly decreased. In addition, page 5 of the model application, response to question 3, states that the margins of safety are not significantly reduced. Considering the information provided in the traveler, the NRC staff could not verify these statements are correct. Provide justifications for these claims. If the staff gains adequate understanding of the other RAIs associated to this project, the answer to this RAI may be evident.

Response to RAI 7 The question is referring to the No Significant Hazards Consideration Analysis in the TSTF-576 model application. The No Significant Hazards Consideration Analysis is provided to support the NRC's determination under 10 CFR 50.92, "Issuance of amendment," of whether the license amendment may be issued prior to any requested hearing on the licensing action. Paragraph (c),

which states, "The Commission may make a final determination, under the procedures in

§ 50.91, that a proposed amendment to an operating license or a combined license for a facility or reactor licensed under §§ 50.21(b) or 50.22, or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety."

The first standard in 10 CFR 50.92 states, "Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?" As stated in the "Applicable Safety Analysis" Bases of the affected TS, the overpressure protection system must accommodate the most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position). The proposed change does not affect the MSIVs and would have no effect on the probability of the MSIVs closing or generation of a reactor scram signal on MSIV position.

Therefore, the probability of the event is unaffected. The consequences of the accident are based on the peak reactor pressure vessel pressure. Both the current and proposed TS ensure the overpressure Safety Limit is not exceeded. As a result, the consequences of the accident are not changed.

The third standard in 10 CFR 50.92 states, "Does the proposed amendment involve a significant reduction in a margin of safety?" NRC Regulatory Issue Summary 2001-22, "Attributes of a Proposed No Significant Hazards Consideration," provides guidance on addressing the third standard. It states, "Licensees should identify the safety margins that may be affected by the proposed change and review the conservatism in the evaluation and analysis methods that are used to demonstrate compliance with regulatory and licensing requirements." The safety margin of interest in the proposed change is the overpressure Safety Limit. The conservatisms in the evaluation and the analysis are described in the NRC-approved methods for each licensee, which are not altered by the proposed change.

NRC Regulatory Issue Summary 2001-22 also states, "If a change does not exceed or alter a design basis or safety limit (i.e., the controlling numerical value for a parameter established in 19

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" the UFSAR or the license) it does not significantly reduce the margin of safety." The only design basis limit or safety limit related to the proposed change is the overpressure Safety Limit.

Both the current TS and the proposed TS are based on ensuring that the overpressure safety limit is protected. Therefore, the margin of safety is not affected.

20

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" Response to NRC Additional Questions Provided After the December 7, 2020 Teleconference The NRC staff's additional questions were unnumbered. For convenience, the questions have been numbered by continuing the sequence from the RAI.

8. In the example scenarios presented, there were two changes made to the Cycle N reload analysis (RLA) based on the test results from Cycle N-1. These changes were to the setpoint tolerance (i.e., increasing from +4% to either +5% or +6%) and use of the 11th valve (which was originally assumed out of service). Describe how these specific changes would be addressed under 10 CFR 50.59 and whether prior NRC approval would be required in order to make these specific changes to the modelling assumptions in the RLA. Based on the guidance in Section 3.4, Departure from a Method of Evaluation Described in the FSAR (as Updated), of NEI 96-07, Revision 1, it appears to the NRC staff that crediting use of the 11th valve would be a non-conservative change and require prior NRC approval. It was also discussed during the meeting that licensees may decide (based on new information) to reduce the setpoint tolerance assumed (i.e., lower from +6% back down to +4%). Would this change require prior NRC approval based on the 10 CFR 50.59 criteria?

Response to RAI 8 Changes to setpoint tolerances and/or number of valves assumed out of service would be considered changes in input parameters, as defined in Section 3.8, "Input Parameters," of NEI 96-07 Revision 1. This is because they describe physical characteristics of plant SSCs (the S/RVs). The NRC-approved methodologies used to perform the cycle-specific RLA overpressure analyses do not require the assumption that one S/RV is out of service. Thus, changes to setpoint tolerances and/or consideration of S/RVs in or out of service are considered changes to the facility and would be evaluated under the other criteria of 10 CFR 50.59(c)(2).

Such changes would not be considered a methodology change for the purposes of Section 3.4 of NEI 96-07.

Similarly, if the S/RVs consistently open closer to the setpoint than represented by the setpoint tolerances used as inputs to the overpressure RLA (perhaps because of installing S/RVs of a new design), then this information may be utilized by the utility as a basis for reducing the assumed tolerances. Such an update would be consistent with other updates to input parameters to the safety analyses performed by the licensee.

21

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements"

9. The basis for the elimination of the TS requiring action if a specific number of SRVs are inoperable has not been presented considering the current TS and Bases.

Response to RAI 9 See response to RAI 6.

10. If SRV test data results in the RLA being updated (with a higher S/RV opening pressure or use of an SRV that was originally assumed out of service, etc.), would any changes be required to a licensee's FSAR? The plants were licensed assuming that the valves would respond as designed, however, this is not the case. If a licensee has S/RVs that frequently open above the ASME +3% tolerance, does a licensees FSAR need to be updated to acknowledge that the valves do not operate as intended? Would the FSAR need to be updated to reflect the change in treatment of the S/RVs afforded by the proposed TS change?

a) In Scenario 1 (slide 4), it states that the Cycle N RLA have been validated as conservative based directly on the Cycle N-1 test results. As discussed during the meeting, the way this is stated in the presentation places too much emphasis on the previous cycle test results. In this case, the Cycle N RLA may be acceptable provided the RLA assumptions (i.e., number of valves opening at a given setpoint, number assumed out of service, etc.) already consider historical data (i.e., more than just the previous cycles test results which could have been good, typical or bad). If the traveler is approved, should there be a condition upon implementation that licensees need to provide their initial RLA assumptions regarding number of valves and opening setpoints along with historical valve test results to demonstrate that what they are using is realistically conservative? The methodology presented in paragraph 5 of the TSTF presentation for the 12/7/20 meeting is not well enough defined. How does such an open-ended TS provide compliance with the regulations?

b) More than just the previous cycle of the historical valve lift data should be considered in some manner that retains adequate conservatism.

Response to RAI 10 The FSARs for some licensees may contain documentation of the values of the standard set of inputs utilized in performing the overpressure analysis. Thus, for these plants, updates to the licensee's FSAR would be necessary to reflect any updated values utilized as input to the safety analyses.

Note that the +/-3% tolerance (or the plant's current design basis) would still be applied to the ASME O&M Code testing. Any FSAR descriptions of S/RV individual design and testing would not be affected by the proposed change.

There is no fundamental methodology change being proposed by the traveler. The traveler provides licensees with the ability to select parameters which are representative of plant operation as inputs to the overpressure analysis performed using the NRC-approved methodology. Therefore, an FSAR update may not be required. As mentioned above, if a 22

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" utility's FSAR currently describes the input parameters used in the reload analyses including the S/RV setpoints and tolerances, then the FSAR may require revision.

Response to RAI 10.a The example scenario is not intended to suggest that a licensee would only consider the most recent as-found test results when establishing the overpressure analysis inputs. The licensee would have a documented basis for the analysis inputs prior to performing the Cycle N-1 as-found tests. In the example scenarios, the new as-found test data either supports or doesn't support the existing basis for the RLA inputs, which is the distinguishing difference between the scenarios.

Licensees are responsible for establishing and maintaining the basis for realistically conservative input parameters to the safety analyses and changes to those input parameters are controlled under 10 CFR 50.59. The significant alteration being accomplished by the proposed change is to base the inputs to the overpressure analysis on the actual performance of the S/RVs rather than on the fixed values in the TS, which in some cases have been shown to not be representative of plant behavior. As discussed in other responses, the proposed change does not alter the NRC-approved methodology for performing overpressure analyses. The NRC does not control the licensee's selection of RLA input assumptions except as provided by 10 CFR 50.59.

Response to RAI 10.b Under the proposed change, older historical valve data will not stop being considered once new data is obtained. As additional data is collected using as-found testing, it will be added as an additional set of information applicable when determining the input assumptions to be used in future cycles, and to be used to check whether the currently operating cycle's RLA inputs are appropriate (i.e., if new data becomes available indicating that valve performance has changed relative to historic data, this needs to be considered).

23

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements"

11. How does the proposed method comply with 10 CFR 50.36(c)(1)(ii)(A)? 10 CFR 50.36 requires in (c)(1)(ii)(A): Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions.

Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. (emphasis added).

Response to RAI 11 The regulatory history of the application of the 10 CFR 50.36 makes it clear that the S/RV lift setpoints fall under 10 CFR 50.36(c)(2), "Limiting Condition for Operation," (LCO), not 10 CFR 50.36(c)(1)(ii)(A), "Limiting Safety System Settings."

The LSSS description was published the December 17, 1968, revision to 10 CFR 50.36, "Technical Specifications," and 10 CFR 50.36(c)(1)(ii)(A) has not changed since that revision other than record retention requirements revised in 1988. The statements of consideration for the 1968 rule referenced a "Guide to Content of Technical Specifications for Nuclear Reactors" published in parallel with the rule for use with the revised system of technical specifications.

That guide referenced the TS for two plants, San Onofre and Haddam Neck, as examples of TS that were developed under the new system of TS. The San Onofre TS are not publicly available, but the Haddam Neck TS are available on microfiche at the address 50213-207. The Haddam Neck TS has a section "Maximum Safety Settings" which uses the same definition as the 10 CFR 50.36 "Limiting Safety System Settings." It lists seven reactor trip signals described as, "trip settings for instruments monitoring reactor power and reactor coolant pressure, temperature, and flow." The pressurizer code safety valve, which serves the same purpose as an S/RV, is described in LCO 3.3, and is not included in the "Maximum Safety Settings."

The NRC's earlier STS for BWR plants, NUREG-0123, with revisions published between 1976 and 1980, Section 2.2, "Limiting Safety System Settings," stated, "The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1." Table 2.2.1-1 did not list S/RV setpoints as an LSSS.

The November 1987 Topical Report, NEDO-31466, "Technical Specifications Screening Criteria Application and Risk Assessment," published by the BWROG as part of the NRC's Technical Specifications Improvement Project (TSIP), determined that the S/RV LCO satisfied Criterion 3 of the NRC's interim policy statement on technical specifications improvement (52FR3788),

which was later added to the regulations as 10 CFR 50.36(c)(2)(ii), Criterion 3. The topical report concluded that the S/RV requirements should be retained as an LCO in the TS. This conclusion was echoed in the NRC's May 9, 1988 letter to the Owners Group chairmen titled, "NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups' Application of the Commission's Interim Policy Statement Criteria to Standard Technical Specifications."

The BWROG's proposed improved STS, NEDC-31681, "Improved BWR Technical Specifications," April 1989, Volume 4, describes changes to the then-current BWR STS (NUREG-0123, 1980) to create the proposed improved STS. NEDC-31681 stated that STS 2.2, "Limiting Safety System Settings," was combined with LCO 3.3.1.1, "RPS Instrumentation," to 24

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" consolidate the RPS requirements into one section. The S/RV setpoints were not identified as an LSSS.

The NRC's improved STS for BWR/4 and BWR/6 plants, NUREG-1433 and NUREG-1434 published in 1992, did not include an LSSS section. The Reactor Protection System Instrumentation Bases stated:

The protection and monitoring functions of the RPS have been designed to ensure safe operation of the reactor. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance. The LSSS are defined in this Specification as the Allowable Values, which, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits, including Safety Limits (SLs) during Design Basis Accidents (DBAs).

There were no other references to limiting safety system settings in the BWR STS Bases, supporting the conclusion that the LSSS were incorporated into the RPS Instrumentation LCO.

Between 2005 and 2010, the NRC and the TSTF discussed TSTF-493, "Clarify Application of Setpoint Methodology for LSSS Functions." TSTF-493, Revision 4, was approved by the NRC on May 11, 2010 (75FRN26294). The traveler addressed limiting safety system settings (LSSSs) assessed during periodic testing and calibration of instrumentation that may have an adverse effect on equipment operability. The approved change revised the Bases to discuss the LSSS.

Only LCOs in Section 3.3, "Instrumentation," were revised, including the BWR/6, "Relief and Low-Low-Set (LLS) Instrumentation," Bases which states:

The safety/relief valves (S/RVs) prevent overpressurization of the nuclear steam system.

Instrumentation is provided to support two modes of S/RV operation - the relief function (all valves) and the LLS function (selected valves). Refer to LCO 3.4.4, "Safety/Relief Valves (S/RVs)," and LCO 3.6.1.6, "Low-Low Set (LLS) Safety/Relief Valves (S/RVs),"

for Applicability Bases for additional information of these modes of S/RV operation.

This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the Safety/Relief valve instrumentation, as well as LCOs on other reactor system parameters, and equipment performance.

The BWR/4 STS does not credit the relief mode of S/RVs for overpressure protection. The S/RV specification (BWR/4 3.4.3 and BWR/6 3.4.4) Bases do not discuss LSSS or state that the S/RV setpoints are LSSS. This supports the position that only instrumentation setpoints are LSSS, not mechanical valve lift setpoints.

Based on the regulatory history, the industry concludes that certain instrumentation may be governed by both the LCO and LSSS requirements, including the relief-mode S/RV setpoints for plants that credit relief-mode S/RVs for overpressure protection, but that non-instrumentation LCOs such as S/RV safety-mode setpoints are not LSSS. Note that the relief-mode setpoints are unaffected by the proposed change and remain in the TS.

However, even if the S/RV safety-mode mechanical lift settings were determined to be LSSSs, the proposed change would satisfy the requirements of 10 CFR 50.36. 10 CFR 50.36 requires 25

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" categories of items to be in the TS, but there is no regulatory requirement for specific values to be in the TS. The NRC has repeatedly permitted LCO and Surveillance Requirement (SR) limits and LSSS setpoint values to be placed under licensee control and concluded that 10 CFR 50.36 was met.

In Generic Letter 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," dated October 4, 1988, the NRC encouraged licensees to submit license amendments to remove plant-specific parameter limits from the TS to be controlled by the licensee under 10 CFR 50.59. In NUREG-1433 for example, there are seven LCOs and three SRs that reference limits in the COLR. In the Westinghouse plant STS, NUREG-1431, the reactor core Safety Limit value, as well as the Overtemperature T and Overpower T parameters are in the COLR. The Overtemperature T and the Overpower T limits ensure the Reactor Core Safety Limits are not exceeded. In all these cases, the staff concluded that the requirements of 10 CFR 50.36 are met.

As discussed above, the LSSS limits were relocated to TS Section 3.3, "Instrumentation," during development of the standard TS. The STS requirements were clarified with the NRC approval of TSTF-493, Rev. 4, "Clarify Application of Setpoint Methodology for LSSS Functions," on May 11, 2010. TSTF-493 included an option to relocate most of the Section 3.3, "Instrumentation,"

setpoints (including the setpoints that are described as LSSS in the TS Bases) to licensee control under the Setpoint Control Program.

The NRC's approved Design Control Document for the Westinghouse Advanced Passive 1000 (AP1000) plant, Chapter 16, "Technical Specifications," Specification 5.5.14, "Setpoint Program (SP)," states, "The Setpoint Program (SP) implements the regulatory requirement of 10 CFR 50.36(c)(1)(ii)(A) that technical specifications will include items in the category of limiting safety system settings (LSSS), which are settings for automatic protective devices related to those variables having significant safety functions." The LSSS setpoints do not appear in the AP1000 TS and the SP allows the licensee to establish the LSSS setpoints under 10 CFR50.59. It states, "The SP shall establish a document containing the current value of the specified NTS, AFT, and ALT for each Technical Specification required automatic protection instrumentation function and references to the calculation documentation. Changes to this document shall be governed by the regulatory requirement of 10 CFR 50.59."

The NRC's approved Design Control Document for the NuScale Nuclear Power Plant and the General Electric ESBWR also contain a Setpoint Program with the same provisions for licensee control of LSSS setpoint values.

The NRC's approved Design Control Document for the Korean APR1400 contains a similar Setpoint Control Program. Of note, the Reviewer's Note describing the scope of the program states, "Settings associated with safety relief valves are excluded. The performance of these components is already controlled (i.e., trended with as-left and as-found limits) under the ASME Code for Operation and Maintenance of Nuclear Power Plants testing program."

In summary, the industry concludes that certain instrumentation may be governed by both the LCO and LSSS requirements, but that the S/RV safety mode setpoints are not LSSS. Even if the 26

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements" S/RV safety mode setpoints were LSSS, it would not prohibit the proposed changes to the S/RV specification.

27

TSTF Response to NRC Questions on TSTF-576, Revision 0, "Revise Safety/Relief Valve Requirements"

12. Which of the example scenarios would have resulted in reporting to the NRC? Would it just be those where the RLA exceeded the safety limit? Would any of the example scenarios be considered as the plant being in an unanalyzed condition that significantly degrades plant safety?

Response to RAI 12 The Technical Specifications do not define reportability requirements and reportability is not discussed in TSTF-576. It is the responsibility of a licensee to evaluate an operational event against the requirements of 10 CFR 50.72 and 10 CFR 50.73 considering the guidance in NUREG-1022, "Event Report Guidelines," and their procedures. As stated in the 1983 Statements of Consideration for 10 CFR 50.72 and 50.73, "The Commission recognizes that the licensee may use engineering judgment and experience to determine whether an unanalyzed condition existed." As such, a definitive statement cannot be made regarding which of the example scenarios would be reportable as an unanalyzed condition that significantly degrades plant safety.

28

TSTF-576, Rev. 1 Technical Specifications Task Force Improved Standard Technical Specifications Change Traveler Revise Safety/Relief Valve Requirements NUREGs Affected: 1430 1431 1432 1433 1434 2194 Classification: 1) Technical Change Recommended for CLIIP?: Yes Correction or Improvement: Improvement NRC Fee Status: Not Exempt Benefit: Increases Equipment Operability Changes Marked on ISTS Rev 5.0 PWROG RISD & PA (if applicable): N/A N/A See attached.

Revision History TSTF Revision 0 Revision Status: Closed Revision Proposed by: BWROG LC Revision

Description:

Original Issue On August 27, 2019, the TSTF provided for NRC comment draft traveler TSTF-576, "Revise Safety/Relief Valve Requirements." A presubmittal meeting was held on September 12, 2019 and on October 21, 2019 the TSTF provided a revised draft. An additional presubmittal meeting was held on December 2, 2019.

On December 13, 2019, the TSTF submitted for NRC review Revision 0 of TSTF-576 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19347A726).

Owners Group Review Information Date Originated by OG: 12-Jul-19 Owners Group Comments Presubmittal meeting held September 12, 2019. Revised traveler distributed to BWROG on October 7.

Owners Group Resolution: Approved Date: 02-Aug-19 TSTF Review Information TSTF Received Date: 03-Dec-19 Date Distributed for Review 03-Dec-19 TSTF Comments:

A presubmittal meeting was held with the NRC on September 12, 2019. A revised draft was developed and submitted to the NRC on October 21. A presubmittal teleconference was held on December 2. The traveler was finalized addressing the NRC comments.

TSTF Resolution: Approved Date: 12-Dec-19 NRC Review Information NRC Received Date: 13-Dec-19 23-Jun-21 Copyright(C) 2021, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-576, Rev. 1 TSTF Revision 0 Revision Status: Closed NRC Comments:

A presubmittal meeting was held with the NRC on September 12, 2019. A revised draft was developed and submitted to the NRC on October 21. A presubmittal teleconference was held on December 2. The traveler was finalized addressing the NRC comments and submitted December 13, 2019.

The traveler was revised to reflect responses to the NRC's May 11, 2020 Request for Additional Information.

Final Resolution: Superceded by Revision TSTF Revision 1 Revision Status: Active Revision Proposed by: BWROG Revision

Description:

On May 11, 2020, the NRC provided a Request for Additional Information (RAI) regarding TSTF-576 (ADAMS Accession Number ML19351D783).

On August 11, 2020, the TSTF provided a draft RAI response and revised traveler for NRC comment. A teleconference to discuss the NRC's comments as held on October 13, 2020, followed by a teleconference audit on December 7, 2020.

On February 1, 2021, a revised RAI response and traveler were provided to the NRC for comment. The NRC provided comments on April 12, 2021 and a teleconference to discuss the NRC comments was held on May 13, 2021.

TSTF-576 was also revised to be based in the completed but not yet published Revision 5 of the Standard Technical Specifications. This did not result in any changes.

The significant changes in Revision 1 are:

1.Searched the ITS for "Safety/Relief" and "S/RV" and replaced statements that the S/RVs are used to ensure that the peak vessel pressure is maintained within the applicable ASME Code limits with a statement that the Overpressure Protection System is used to ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

2.Expanded justification of elimination of manual actuation test and added statement that licensees adopting the traveler have an adequate FME program.

3.Removed the BWR/4 NUREG addition of an SR on relief mode S/RVs that is only applicable to Dresden and Quad Cities. Added an acceptable variation for those plants to retain their existing SR and Actions.

4.Restored an LCO Bases description of the ASME Code requirements.

5.Revised the Required Action A.1 removal justification based on a review the S/RV TS for all BWRs.

6.Added a paragraph to the Bases regarding the removal of the SR 3.4.3.1 low tolerance value.

7.Revised and expanded the justification for removal of the manual actuation SR based on a review the S/RV TS for all BWRs.

8.Many editorial revisions and improvements.

23-Jun-21 Copyright(C) 2021, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-576, Rev. 1 TSTF Revision 1 Revision Status: Active Owners Group Review Information Date Originated by OG: 21-May-21 Owners Group Comments (No Comments)

Owners Group Resolution: Approved Date: 23-Jun-21 TSTF Review Information TSTF Received Date: 07-Jun-21 Date Distributed for Review 07-Jun-21 TSTF Comments:

(No Comments)

TSTF Resolution: Approved Date: 23-Jun-21 NRC Review Information NRC Received Date: 23-Jun-21 NRC Comments:

First draft provided to NRC on 8/11/2020. The draft was revised based on NRC comments on the draft RAI response.

An audit was held on December 17, 2020. The NRC provided additional questions following the audit.

On February 1, 2021, a revised RAI response and traveler were provided to the NRC for comment. The NRC provided comments on April 12, 2021 and a teleconference to discuss the NRC comments was held on May 13, 2021 Affected Technical Specifications S/A 2.1.2 Bases RCS Pressure SL S/A 3.1.4 Bases Control Rod Scram Times S/A 3.3.1.1 Bases RPS Instrumentation S/A 3.3.4.2 Bases ATWS-RPT Instrumentation S/A 3.4.12 Reactor Steam Dome Pressure NUREG(s)- 1433 1434 Only Bkgnd 3.6.1.6 Bases LLS Valves SR 3.3.6.3.7 Bases LLS Instrumentation NUREG(s)- 1433 Only 3.4.3 S/RVs NUREG(s)- 1433 Only Change

Description:

Specification renamed Overpressure Protection System 23-Jun-21 Copyright(C) 2021, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-576, Rev. 1 3.4.3 Bases S/RVs NUREG(s)- 1433 Only Change

Description:

Specification renamed Overpressure Protection System Bkgnd 3.4.3 Bases S/RVs NUREG(s)- 1433 Only S/A 3.4.3 Bases S/RVs NUREG(s)- 1433 Only LCO 3.4.3 S/RVs NUREG(s)- 1433 Only LCO 3.4.3 Bases S/RVs NUREG(s)- 1433 Only Appl. 3.4.3 Bases S/RVs NUREG(s)- 1433 Only Action 3.4.3.A S/RVs NUREG(s)- 1433 Only Change

Description:

Deleted Action 3.4.3.A Bases S/RVs NUREG(s)- 1433 Only Change

Description:

Deleted Action 3.4.3.B S/RVs NUREG(s)- 1433 Only Change

Description:

Deleted Action 3.4.3.B Bases S/RVs NUREG(s)- 1433 Only Change

Description:

Deleted Action 3.4.3.C S/RVs NUREG(s)- 1433 Only Change

Description:

Revised and renamed "A" Action 3.4.3.C Bases S/RVs NUREG(s)- 1433 Only Change

Description:

Revised and renamed "A" SR 3.4.3.1 S/RVs NUREG(s)- 1433 Only SR 3.4.3.1 Bases S/RVs NUREG(s)- 1433 Only SR 3.4.3.2 S/RVs NUREG(s)- 1433 Only Change

Description:

Deleted SR 3.4.3.2 Bases S/RVs NUREG(s)- 1433 Only Change

Description:

Deleted Ref. 3.4.3 Bases S/RVs NUREG(s)- 1433 Only S/A 3.4.11 Bases Reactor Steam Dome Pressure NUREG(s)- 1433 Only Bkgnd 3.3.6.5 Bases Relief and LLS Instrumentation NUREG(s)- 1434 Only 3.4.4 S/RVs NUREG(s)- 1434 Only Change

Description:

Specification renamed Overpressure Protection System 3.4.4 Bases S/RVs NUREG(s)- 1434 Only Change

Description:

Specification renamed Overpressure Protection System 23-Jun-21 Copyright(C) 2021, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-576, Rev. 1 Bkgnd 3.4.4 Bases S/RVs NUREG(s)- 1434 Only S/A 3.4.4 Bases S/RVs NUREG(s)- 1434 Only LCO 3.4.4 S/RVs NUREG(s)- 1434 Only LCO 3.4.4 Bases S/RVs NUREG(s)- 1434 Only Appl. 3.4.4 Bases S/RVs NUREG(s)- 1434 Only Action 3.4.4.A S/RVs NUREG(s)- 1434 Only Change

Description:

Deleted Action 3.4.4.A Bases S/RVs NUREG(s)- 1434 Only Change

Description:

Deleted Action 3.4.4.B S/RVs NUREG(s)- 1434 Only Change

Description:

Deleted Action 3.4.4.B Bases S/RVs NUREG(s)- 1434 Only Change

Description:

Deleted Action 3.4.4.C S/RVs NUREG(s)- 1434 Only Change

Description:

Renamed A Action 3.4.4.C Bases S/RVs NUREG(s)- 1434 Only Change

Description:

Renamed A SR 3.4.4.1 S/RVs NUREG(s)- 1434 Only SR 3.4.4.1 Bases S/RVs NUREG(s)- 1434 Only SR 3.4.4.2 S/RVs NUREG(s)- 1434 Only SR 3.4.4.2 Bases S/RVs NUREG(s)- 1434 Only SR 3.4.4.3 S/RVs NUREG(s)- 1434 Only Change

Description:

Deleted SR 3.4.4.3 Bases S/RVs NUREG(s)- 1434 Only Change

Description:

Deleted Ref. 3.4.4 Bases S/RVs NUREG(s)- 1434 Only 23-Jun-21 Copyright(C) 2021, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-576, Rev. 1

1.

SUMMARY

DESCRIPTION The proposed change revises the Safety/Relief Valve (S/RV) Technical Specifications (TS) to align the overpressure protection requirements with the safety limits and the regulations. The proposed change modifies NUREG-1433, "Standard Technical Specifications, General Electric BWR/4 Plants," and NUREG-1434, "Standard Technical Specifications, General Electric BWR/6 Plants."1

2. DETAILED DESCRIPTION 2.1. System Design and Operation The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requires the reactor pressure vessel to be protected from overpressure during upset conditions by self-actuated safety valves. The overpressure protection requirements dictate the size and number of S/RVs that are needed such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB) under the most severe transients. Section 5.2.2, "Overpressure Protection," of NUREG-0800, "Standard Review Plan,"

describes the typical requirements for the overpressure protection system for boiling water reactor (BWR) plants.

Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix A, "General Design Criteria," (GDC), criterion 15 "Reactor coolant system design," states, "The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences." While many of the operating plants are not committed to the Appendix A GDC, most plants are committed to a similar design requirement as described in their Updated Final Safety Analysis Report (UFSAR).

The overpressure protection system for a BWR utilizes the S/RVs. The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. The S/RVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the spring loaded disk or pilot valve opens when steam pressure overcomes the spring force holding the valve or pilot valve closed. For S/RVs with pilot valves, opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve. In the relief mode of operation, pneumatic pressure is used to open the valve, initiated by switches located in the control room or by pressure-sensing instrumentation. Some plants credit a percentage of the total installed S/RV capacity operating via the relief mode for overpressure protection, as permitted by the ASME Code.

1 NUREG 1433 is based on the BWR/4 plant design, but is also representative of the BWR/2, BWR/3, and, in this case, BWR/5 designs. NUREG 1434 is based on the BWR/6 plant design.

Page 1

TSTF-576, Rev. 1 2.1.1. S/RV Inservice Testing The S/RVs are tested in accordance with the Inservice Testing (IST) Program, as required by 10 CFR 50.55a(f). Periodic testing is described in Appendix I of the ASME Operations and Maintenance (OM) Code, "Class 1 Main Steam Pressure Relief Valves with Auxiliary Actuating Devices," Section I-3300, "Periodic Testing." This testing is performed during a plant shutdown and aspects are performed as a bench test at nominal operating temperatures and pressures. The inservice test verifies each S/RV opens within the required "as-found" tolerance around the setpoint.

Safety/Relief Valve nominal setpoints, as-left tolerance limits, and as-found tolerance limits that appear in the current TS are also established and controlled by the ASME OM Code. ASME OM Code Appendix I Section I-1310(e), states, "The Owner, based upon system and valve design basics or technical specification, shall establish and document acceptance criteria for tests required by this Mandatory Appendix." The 2015 edition of the OM Code, section I-1320, "Test Frequencies, Class 1 pressure Relief Valves," paragraph (c), "Requirements for testing additional valves," states, "Additional valves shall be tested in accordance with the following requirements:

(1) For each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of either the plus/minus tolerance limit of the Owner-established set-pressure acceptance criteria of sub-para. I-1310(e) or +/- 3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group." Other editions of the OM Code have similar requirements. Therefore, additional testing is required by the OM Code if an S/RV fails to open within established acceptance criteria (the owner specified limits or +/-3%) or the as-found tolerance established in the TS.

The ASME Code permits testing 20% of the S/RVs each cycle prior to startup, with the test population expanded if failures are found. Alternatively, all of the S/RVs or pilot valves may be removed and replaced, and the as-found testing is performed within one year after removal.

Following testing, the S/RVs or pilot valves are refurbished, tested, and certified for use. The valves are set to the "as-left" tolerance, which is typically narrower than the as-found criteria to allow for drift during the period of operation.

If an S/RV fails to open within the IST tolerance during as-found testing, the failure is entered into the Corrective Action Program and, according to licensee procedures, evaluated, corrected, and tracked. The extent of condition is also evaluated. Depending on the nature of the failure, the extent of condition could include an evaluation of the ability of the S/RVs to perform their function in the current cycle.

As an example of evaluation of S/RV performance under the Corrective Action Program, in 2016 Southern Company discovered unexpected damage during testing of the S/RVs for Plant Hatch Unit 1. After examination, it was determined that the damage was similar to damage reported in a previous 10 CFR Part 21 report. Extensive extent of condition evaluations were performed on Unit 1 and Unit 2 (see NRC Reactive Inspection Report 05000321/2016009 dated June 10, 2016), which determined the Hatch S/RVs were suspectable to fretting as described in the 10 CFR Part 21 report. As a result, in May of 2016 Southern Company performed a mid-cycle outage on Plant Hatch Unit 2 to replace all eleven S/RVs and to inspect the main valve internals.

Page 2

TSTF-576, Rev. 1 2.2. Current Technical Specifications Requirements In addition to the ASME Code requirements, the current TS contain multiple specifications that govern the S/RVs depending on the function they are fulfilling.

• Safety Limit 2.1.2, "Reactor Coolant System Pressure SL," states, "Reactor steam dome pressure shall be 1325 psig." The pressure limit is plant specific. The S/RVs are credited for meeting this safety limit. Safety Limit 2.1.2 limits the reactor steam dome pressure to the lowest transient overpressure allowed in order to ensure the maximum transient pressure allowable in the RCS pressure vessel is less than the ASME Code, Section III, limit of 110%

of design pressure.

• BWR/4 and BWR/6 TS 3.5.1, "ECCS - Operating," requires the Automatic Depressurization System (ADS) to be operable, which uses the S/RVs in the relief mode. The ADS is designed to provide depressurization of the reactor pressure vessel (RPV) during a small break Loss of Coolant Accident (LOCA) if high pressure core injection (BWR/4) or high pressure core spray (BWR/6) fails or is unable to maintain the required water level in the RPV. ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure Emergency Core Cooling System (ECCS) subsystems, so that these subsystems can provide coolant inventory makeup.

• BWR/4 and BWR/6 TS 3.6.1.6, "Low-Low Set (LLS) Valves," requires the S/RVs operating in relief mode to be operable. In the LLS mode, a subset of the S/RVs are signaled to open at a lower pressure than the relief or safety mode pressure setpoints and to stay open longer, so that reopening more than one S/RV is prevented on subsequent actuations. The LLS function prevents excessive short duration S/RV cycles with valve actuation at the relief setpoint.

• BWR/4 TS 3.3.6.3, "Low-Low Set (LLS) Instrumentation," and BWR/6 TS 3.3.6.5, "Relief and Low-low Set (LLS) Instrumentation," provide instrumentation requirements that support the S/RVs in the LLS mode of operation. For plants that credit S/RVs in relief mode to prevent overpressurization, the LLS Instrumentation TS also provide the instrumentation requirements to support that function.

• BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4, both titled, "Safety/Relief Valves," require the S/RVs to prevent RCPB overpressurization. For most plants, the most severe pressurization transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position). For most BWR/2, BWR/3, BWR/4, and BWR/5 plants, the S/RVs in the safety mode ensure the Safety Limit is not exceeded during normal operation and Anticipated Operational Occurrences (AOOs). For BWR/6 plants and two non-BWR/6 plants (Dresden 2 and 3 and Quad Cities 1 and 2), some S/RVs in relief mode in addition to the S/RVs in safety mode are credited to ensure the ASME Code overprotection limit is protected.

The BWR/4 TS 3.4.3 S/RV LCO typically states, "The safety function of XX S/RVs shall be operable," with the required number of S/RVs (XX) corresponding to the minimum number needed to accommodate the limiting pressure transient without exceeding the Safety Limit Page 3

TSTF-576, Rev. 1 using only the safety mode of operation. The LCO may specify fewer S/RVs than are installed in the plant.

The BWR/6 TS 3.4.4 LCO (which is also applicable to two non-BWR/6 plants) requires the safety function of XX S/RVs and the relief function of YY additional S/RVs to be operable.

The required number of S/RVs in the safety mode (XX) and relief mode (YY) varies by plant and may specify fewer S/RVs than are installed.

BWR/4 Surveillance Requirement (SR) 3.4.3.1 and BWR/6 SR 3.4.4.1 require verification of the safety function lift setpoints of the required S/RVs. These SRs reflect the performance of the ASME Code inservice testing and state the number of valves required to open within a specified tolerance (typically 3%) of the given setpoint. The SRs also specify the as-left tolerance (typically 1%) after testing.

BWR/6 SR 3.4.4.2 requires verification that each relief function S/RV actuates on an actual or simulated automatic initiation signal. The two Non-BWR/6 plants that credit the S/RV relief mode have a similar SR.

BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3 verify that each S/RV opens when manually actuated.

2.3. Reason for the Proposed Change The S/RV LCO is written in terms of individual valves, but the specified safety function is based on the combined pressure relieving capacity of a group of the S/RVs. The failure of some valves to open within the SR tolerance typically would not result in the inability of the S/RVs as a group to perform the specified safety function. Therefore, the LCO should be revised to align with the specified safety function.

Testing of the safety mode of each S/RV is required by the IST Program, which is required by 10 CFR 50.55a(f). It is unnecessary to duplicate this regulatory requirement in the TS when the result of any individual valve test is not required to meet the specified safety function of the system.

A review of Licensee Event Reports over the last ten years found over forty events in which S/RVs failed to lift within the SR lift pressure tolerance when bench tested. In all cases in which the SR was not met due to setpoint drift, the Licensee Event Reports concluded that the S/RVs as a group would have retained the capability to protect Safety Limit 2.1.2. This represents an unnecessary reporting burden on the licensees for failures that did not affect the ability to perform the specified safety function.

2.4. Description of the Proposed Change The proposed changes are based on the completed but not yet published Revision 5 of the Standard Technical Specifications, NUREG-1433 and NUREG-1434.

The proposed change renames BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4 from "Safety/Relief Valves (S/RVs)" to "Overpressure Protection System (OPS)." This title change requires revision to the Page 4

TSTF-576, Rev. 1 Table of Contents and a reference in the Bases of BWR/4 TS 3.3.6.3, "Low-Low Set (LLS)

Instrumentation," and BWR/6 TS 3.3.6.5, "Relief and Low-low Set (LLS) Instrumentation."

The proposed change revises the S/RV LCO to require the Overpressure Protection System (OPS) to be operable. The LCO Bases describes an operable OPS as being capable of preventing reactor steam dome pressure from exceeding Safety Limit 2.1.2.

BWR/4 LCO 3.4.3 is revised to state (deletions are struck through; insertions are in italics):

The OPS safety function of the [11] S/RVs shall be OPERABLE.

BWR/6 LCO 3.4.4 is revised to state:

The OPS safety function of the [seven] S/RVs shall be OPERABLE, AND The relief function of [seven] additional S/RVs shall be OPERABLE.

BWR/4 SR 3.4.3.1 is revised to state:

Verify the OPS has the capability to prevent reactor steam dome pressure from exceeding Safety Limit 2.1.2.

---

NOTE------------------------------

[2] [required] S/RVs may be changed to a lower setpoint group.

Verify the safety function lift setpoints of the [required] S/RVs are as follows:

Number of Setpoint S/RVs (psig)

[4] [1090 +/- 32.7]

[4] [1100 +/- 33.0]

[3] [1110 +/- 33.3]

Following testing, lift settings shall be within +/- 1%.

BWR/6 SR 3.4.4.1 is revised to state:

Verify the OPS has the capability to prevent reactor steam dome pressure from exceeding Safety Limit 2.1.2.

---

NOTE------------------------------

[2] [required] S/RVs may be changed to a lower setpoint group.

Page 5

TSTF-576, Rev. 1 Verify the safety function lift setpoints of the [required] S/RVs are as follows:

Number of Setpoint S/RVs (psig)

[8] [1165 +/- 34.9]

[6] [1180 +/- 35.4]

[6] [1190 +/- 35.7]

Following testing, lift settings shall be within +/- 1%.

The current frequency has three options: In accordance with the Inservice Testing Program, [18]

months, or in accordance with the Surveillance Frequency Control Program. The [18] month and Surveillance Frequency Control Program options are deleted.

BWR/6 SR 3.4.4.2 is revised to state:

Verify each [required] relief function safety/relief valve acting in the relief mode S/RV actuates on an actual or simulated automatic initiation signal.

BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3, which state, "Verify each [required] S/RV opens when manually actuated," are deleted.

The changes to the LCO and SRs result in changes to the TS Actions.

BWR/4 and BWR/6 Condition A, "One [or two] [required] S/RV[s] inoperable," and "One

[required] S/RV inoperable," respectively, are deleted as the LCO and SRs no longer contain requirements on individual S/RVs.

Condition B, the default action when the Condition A Required Action and associated Completion Time is not met, is no longer required after deletion of Condition A and is removed.

BWR/4 and BWR/6 Condition C, "[Three] or more [required] S/RVs inoperable," and "[Two] or more [required] S/RVs inoperable," respectively, are replaced with a new Condition, "OPS inoperable." The new Condition retains the existing Required Actions to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The TS Bases are revised to reflect the changes to the TS. The Bases of the following specifications state that S/RVs are used to ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

• 2.1.2, Reactor Coolant System (RCS) Pressure SL

• 3.1.4, Control Rod Scram Times

• 3.3.1.1, RPS Instrumentation

• 3.3.4.2, ATWS-RPT Instrumentation Page 6

TSTF-576, Rev. 1

• BWR/4 3.3.6.3, LLS Instrumentation

• BWR/ 3.3.6.5, Relief and LLS Instrumentation

• BWR/4 3.4.11, Reactor Steam Dome Pressure

• BWR/6 3.4.12, Reactor Steam Dome Pressure

• 3.6.1.6, LLS Valves These Bases are revised to state that the Overpressure Protection System is used to ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

The regulation at Title 10 of the Code of Federal Regulations (10 CFR), Part 50.36, states, "A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications." A licensee may make changes to the TS Bases without prior NRC review and approval in accordance with the Technical Specifications Bases Control Program.

The proposed TS Bases changes are consistent with the proposed TS changes and provide the purpose for each requirement in the specification consistent with the Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 2, 1993 (58 FR 39132). Therefore, the Bases changes are provided for information and approval of the Bases is not requested.

A model application is attached. The model may be used by licensees desiring to adopt the traveler following NRC approval.

3. TECHNICAL EVALUATION Specification Name Change.

As discussed in Section 2.2, there are several specifications which provide requirements on the S/RVs. It is confusing to title the specification "Safety/Relief Valves," because that name implies it is the only specification that governs the equipment. Just as TS 3.5.1 refers to the "Automatic Depressurization System (ADS)" function of the S/RVs, it is more appropriate to title BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4 "Overpressure Protection System (OPS)," to represent the functional capability required by the specification. Renaming the specifications is consistent with the STS convention that an LCO requires a system to be operable and the LCO Bases describe what is required for the system to capable of performing its specified safety function. The term "overpressure protection system," is not new. The NRC Standard Review Plan (NUREG-0800), Section 5.2.2, is titled, "Overpressure Protection," and many BWR plants have a similar Updated Final Safety Analysis Report (UFSAR) section. In addition, the existing BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4 "Applicable Safety Analysis" section of the Bases begins, "The overpressure protection system must accommodate the most severe pressurization transient." As a result, referring to the S/RV overpressure protection function as the "Overpressure Protection System (OPS)," is a clearer representation of the requirement.

LCO Changes Title 10 of the CFR, Paragraph 50.36(c)(2)(i) states, "Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of Page 7

TSTF-576, Rev. 1 the facility." However, the existing S/RV LCO does not represent the lowest functional capability required for safe operation. The existing S/RV LCO requires each specified valve to be operable, but the specified safety function is based on the combined pressure relieving capacity of the credited S/RVs (which may be less than the installed complement of valves). The failure of a particular valve or valves to open within the SR tolerance may not (and based on historical performance, is unlikely to) result in the inability of the S/RVs as a group to perform the specified safety function. As a result, the existing LCO does not represent the lowest functional capability or performance level of equipment required for safe operation of the facility. Therefore, the LCO is revised to ensure the safety function of providing overpressure protection, which is consistent with 10 CFR 50.36(c)(2)(i).

TSTF-GG-05-02, "Writers Guide for Plant-Specific Improved Technical Specifications,"

(ADAMS Accession No. ML070660229) Section 4.1.4, "Chapter 3 LCO Content," states, "The LCO describes as simply as possible the lowest functional capability or performance levels of equipment required for safe operation of the facility. ... It is acceptable to generically refer to the system, subsystem, component or parameter which is the subject of the LCO and provide the specific scope/boundaries in the Bases." Following this guidance, the BWR/4 TS 3.4.3 and BWR/6 TS 3.4.4 LCOs are revised to require the OPS to be operable. The LCO Bases are revised to state, "The OPS is OPERABLE when it can ensure that the ASME Code limit on peak reactor pressure, as stated in Safety Limit 2.1.2, will be protected using the safety [and relief] mode[s] of the S/RVs. The OPERABILITY of the OPS is only dependent on the ability to relieve excess pressure and maintain reactor pressure below Safety Limit 2.1.2, and may credit less than the full complement of installed S/RVs." The phrase "and relief" is bracketed (i.e., plant-specific) in the BWR/4 TS Bases since it is applicable to only two plants. The phrase is not bracketed in the BWR/6 TS since it is applicable to all BWR/6 plants.

The terms "safety function" and "relief function" are used in the existing BWR/4 LCO 3.4.3 and BWR/6 LCO 3.4.4. However, the TS Bases, "Background" section uses the terms "safety mode" and "relief mode." For example, the Bases state, "The S/RVs can actuate by either of two modes: the safety mode or the relief mode," and "The S/RVs that provide the relief mode are the low-low set (LLS) valves and the Automatic Depressurization System (ADS) valves." The term "safety function" could be easily confused with the term "specified safety function" used in the definition of operability. For clarity and for consistency, the TS and Bases are revised to use the terms "safety mode" and "relief mode." This is an administrative change with no change in intent.

The LCO is revised to no longer specify the number of credited operable S/RVs. As stated previously, the overpressure protection function is provided by the collective action of the credited S/RVs, not individual S/RVs. This change is consistent with the required function and the 10 CFR 50.36 requirement that the LCO represent the lowest functional capability required for safe operation of the facility. The proposed Applicable Safety Analysis Bases describe the number of S/RVs credited in the overpressure analyses.

The BWR/6 LCO requires relief mode of operation for a subset of S/RVs. These plant designs permit crediting some of the pressure relieving capability of electrically operated pressure relief valves in the overpressure analysis. Two non-BWR/6 plants, Dresden 2 and 3 and Quad Cities 1 and 2, also credit electrically operated pressure relief valves and changes are proposed to Page 8

TSTF-576, Rev. 1 accommodate that design. The revision to the LCO to require the OPS to be operable includes a change to the Bases to describe the role of the S/RVs in relief mode. The Applicability Safety Analysis Bases are revised to state that the S/RVs credited for the relief mode are in addition to those credited in the safety mode portion, consistent with the term "additional" which appears in the current LCO. The required relief mode operation of the S/RVs is verified in an SR.

BWR/4 SR 3.4.3.1 and BWR/6 SR 3.4.4.1 Changes Paragraph 10 CFR 50.36(c)(3) states, "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." While the existing test of each S/RV assures the necessary quality of the components is maintained, the testing is duplicative of the IST Program, which is required by regulation (10 CFR 50.55a(f)). The existing SR meets the regulatory guidance that the SR will assure that facility operation will be within safety limits and verifies that the existing LCO is met, but as discussed above, the LCO is not consistent with 10 CFR 50.36(c)(2)(i) because it requires more than the lowest functional capability or performance level of equipment required for safe operation of the facility.

The existing SR verifies each S/RV lifts within the tolerance around the specified setpoint. As previously discussed, the specified safety function is based on the collective capability of the credited S/RVs to relieve pressure, not the ability of each S/RV to lift within a specified tolerance. The LCO requirement that the S/RVs be operable is replaced with a requirement that the Overpressure Protection System be operable. Therefore, the SR is replaced with a requirement to verify the OPS has the capability to prevent reactor steam dome pressure from exceeding Safety Limit 2.1.2.

The SR Bases are revised to discuss the relationship between the SR and the inservice testing of the S/RVs required by the ASME Code. The proposed Bases state:

This Surveillance verifies that the OPS has the capability to prevent the reactor steam dome pressure from exceeding Safety Limit 2.1.2. The testing of the S/RV safety mode lift settings is performed during shutdown, since this is a bench test, to be done in accordance with the INSERVICE TESTING PROGRAM. The measured S/RV mechanical lift pressures determined in accordance with the INSERVICE TESTING PROGRAM are reviewed and compared to the overpressure analysis to verify that the collective performance of the credited S/RVs will ensure Safety Limit 2.1.2 is protected.

Should one or more of the credited S/RVs not actuate within the assumed tolerance during as-found testing, the actual lift values will be used to evaluate the affected previous cycle overpressure analyses to determine whether the Safety Limit was protected. In this case, the SR is met by a combination of testing and calculation. The effect on OPS operability of S/RVs not actuating within the assumed tolerance during as-found testing will be evaluated for the current cycle under the corrective action program using the NRC-approved methods for overpressure and accident analyses.

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TSTF-576, Rev. 1 The description of the OPS in the LCO Bases includes the safety mode of the S/RVs and, when necessary, the relief mode of additional S/RVs. Therefore, when the SR requires verification that the OPS will protect the Safety Limit, it includes the relief mode when credited in the overpressure protection analysis. As a result, it is not necessary to call out S/RVs operating in the relief mode in the SR. However, an SR describes requirements on the relief mode S/RVs.

The proposed change to the SR no longer specifies the number of S/RVs set at each lift setpoint and the as-found tolerance around the setpoint. This information is controlled by the ASME OM Code, which is required to be followed by 10 CFR 50.55a. Appendix I of the ASME OM Code provides S/RV testing requirements, including establishment of setpoints, as-found tolerances, and as-left tolerances. ASME OM Code requirement I-1310(e), states, "The Owner, based upon system and valve design basics or technical specification, shall establish and document acceptance criteria for tests required by this Mandatory Appendix." Therefore, the acceptance criteria will be specified in the licensee-controlled documents.

The 2015 edition of the OM Code, section I-1320, "Test Frequencies, Class 1 pressure Relief Valves," paragraph (c), "Requirements for testing additional valves," states, "Additional valves shall be tested in accordance with the following requirements: (1) For each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of either the plus/minus tolerance limit of the Owners-established set-pressure acceptance criteria of sub-para. I-1310(e) or +/-3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group." Section I-1310(e), "Acceptance Criteria," states, "The Owner, based upon system and valve design basics or technical specification, shall establish and document acceptance criteria for tests required by this mandatory Appendix." Other editions of the OM Code have similar requirements.

Under the proposed change, the S/RV testing acceptance criteria that are based on the technical specifications are removed. Therefore, the Owner-established acceptance criteria applies, or

+/- 3% if the Owner has not established a tolerance. Unless the licensee establishes new acceptance criteria, the S/RV testing and maintenance will continue to be based on +/-3% or the tolerance in the current TS. Under the proposed change, the licensee may justify under 10 CFR 50.59 a representative set of S/RV setpoints and tolerances based on historical plant operation to be used as inputs to the overpressure analysis. Doing so will not affect the ASME OM Code testing acceptance criteria and the required testing and associated maintenance.

Periodic testing of S/RVs will still be performed as required by Appendix I of the ASME OM Code, "Class 1 Main Steam Pressure Relief Valves with Auxiliary Actuating Devices," Section I-3300, "Periodic Testing." Title 10 of the CFR, Part 50, paragraph 55a, "Codes and standards,"

requires licensees to follow the ASME OM Code. The results of the OM Code-required testing will be used to evaluate S/RV performance under the proposed BWR/4 SR 3.4.3.1 and BWR/6 SR 3.4.4.1.

The number of S/RVs required at each setpoint and the as-found and as-left setpoints are not specified in the Bases of the proposed change, as inclusion of that information would inappropriately tie individual valve performance to operability of the OPS.

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TSTF-576, Rev. 1 The proposed change removes the low tolerance about the setpoint from the TS. The low tolerance is not an assumption in the overpressure analysis and not needed to protect Safety Limit 2.1.2. However, the Inservice Testing Program will continue to confirm that the S/RVs open between the lower and upper tolerances about the setpoint.

The S/RV safety mode lift setpoints and tolerances are also inputs to the accident analyses. The licensee may set the number of S/RVs at each lift setpoint, and the as-found and as-left tolerances, as specified by the OM Code, and verify that the overpressure and accident analyses provide acceptable results using NRC-approved methods. The S/RV lift setpoints and tolerances used in accident analyses will be maintained in licensee-controlled documents subject to the 10 CFR 50.59 change controls, similar to other analysis assumptions. Should the as-found testing reveal one or more S/RVs that do not open within the assumed tolerance about the setpoint, the effect on the accident analysis will be evaluated under the Corrective Action Program.

The proposed approach is similar to other SRs that require analysis to evaluate whether the SR is met. For example, BWR/4 SR 3.7.5.1 states, "Verify each [control room AC] subsystem has the capability to remove the assumed heat load." As discussed in the associated Bases, performance of the SR consists of a combination of testing and calculation, just as the proposed S/RV SR will require a combination of testing and calculation to verify the SR is met.

The S/RV relief mode setpoints will continue to be specified in BWR/6 TS 3.3.6.5, "Relief and Low-low Set (LLS) Instrumentation," and in some plant-specific TS not based on the STS.

Under the proposed SR, the results of the Inservice Testing Program individual valve testing will be reviewed to verify that the collective performance of the S/RVs will ensure Safety Limit 2.1.2 is protected. If all of the required S/RVs actuate within the assumed tolerance, the SR is met.

If one or more as-found S/RV lift setpoints are not within the inputs and assumptions of the previous cycle RLA, the previous cycle overpressure RLA is reevaluated using revised inputs considering the as-found test results and the NRC-approved methodology for the licensee. The purpose of the evaluation is to determine if the RLA of the previous cycle bounded the actual plant performance and the SR was met. This reevaluation may be performed by the licensee or a vendor. The reevaluation is performed using the measured S/RV lift settings for tested S/RVs and the upper limits of the ASME Code testing allowance for any S/RVs that were not tested.

Any S/RV that was required to be tested but that could not be tested, or if the results cannot be determined, is assumed to be out-of-service.

The previous cycle RLA reevaluation may result in one of the following outcomes:

i. If the previous cycle RLA reevaluation demonstrates that the calculated overpressure is less than or equal to the RLA calculated peak overpressure for the limiting event (i.e., the RLA results were bounding), then the SR was met and the OPS was operable during the previous cycle. It can be assumed that the current cycle RLA inputs and assumptions contain adequate conservatism to account for the as-found S/RV performance. No further action is required.

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TSTF-576, Rev. 1 ii. If the previous cycle RLA reevaluation does not demonstrate that the calculated overpressure is less than or equal to the RLA calculated peak overpressure for the limiting event (i.e., the RLA results are not bounding), the previous cycle overpressure analysis was not consistent with actual plant performance and the issue will be entered into the Corrective Action Program (as required by 10 CFR 50, Appendix B, Criterion XVII, "Corrective Actions,"). The previous cycle performance will be evaluated for reportability under 10 CFR 50.72 and 10 CFR 50.73.

If the previous cycle RLA reevaluation determines the RLA overpressure analysis did not bound actual plant performance, the current cycle overpressure RLA will be reevaluated as part of the Corrective Action Program extent of condition consideration.

The current cycle RLA reevaluation will use the licensee's NRC-approved methodology but the inputs and assumptions will be updated as needed considering the previous cycle test results.

The nature of the changes to the inputs will depend on the cause of the as-found failures and the similarities or differences between the previous cycles and the current cycle. An evaluation of the cause of the as-found failures may result in changes to the S/RV tolerances or other assumptions in the reevaluation. If it is determined that the OPS is not operable, the TS Actions require a plant shutdown.

Safety/Relief Valves fall under 10 CFR 50.65 (the Maintenance Rule). Licensee Maintenance Rule programs require establishing performance criteria, monitoring and trending performance, determining the cause of failures, and taking corrective action. Those activities are available for NRC inspection.

The Boiling Water Reactor Owners' Group (BWROG) has been working to improve S/RV performance for many years and has trended the performance of problematic two-stage S/RVs.

The BWROG plans to continue to monitor and improve S/RV performance across the BWR fleet.

Safety/Relief Valves that are removed from the plant for testing are refurbished, certified, and reset to within the as-left tolerance prior to reinstallation in accordance with the ASME Code.

In summary, the proposed BWR/4 SR 3.4.3.1 and BWR/6 SR 3.4.4.1 will verify that the Overpressure Protection System, which represents the collective function of the S/RVs, will perform the specified safety function, and will confirm that facility operation will be within the safety limits. The requirements on individual S/RVs will be adequately controlled by 10 CFR 50.55a, the ASME OM Code, 10 CFR 50.65, and 10 CFR 50.59, and do not need to appear in the TS.

BWR/6 SR 3.4.4.2 Changes BWR/6 SR 3.4.4.2 states, "Verify each [required] relief function S/RV actuates on an actual or simulated automatic initiation signal." The SR is revised to refer to the "safety/relief valve acting in the relief mode" instead of the "relief function" as previously discussed. The brackets around the word "required" are removed. Brackets indicate a plant-specific option. The equivalent SR in all four BWR/6 plants contains the word "required." Therefore, the brackets are unnecessary and are removed to make the STS consistent with the plant TS.

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TSTF-576, Rev. 1 BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3 Elimination The existing SRs state, "Verify each [required] S/RV opens when manually actuated." The Bases for the SRs state that a manual actuation is performed to verify that, mechanically, the valve is functioning properly, and no blockage exists in the valve discharge line. This SR is removed from the TS.

The STS SR is not representative of the plant-specific TS. Ten of the thirty BWR units have no equivalent SR. An additional thirteen of the thirty units have different requirements, such as verifying the actuator strokes (eleven units) or verification that the valve is capable of being opened (two units). Only seven of the thirty units contain an SR equivalent to the STS SR.

There is no safety analysis assumption that the safety mode S/RVs will open when manually actuated to limit overpressure. As a result, the ability to open manually is not required to demonstrate that the valve is functioning properly. This is supported by the existing SR Bases statement, "If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the S/RV is considered operable." TS SR 3.0.1 states that when an SR is not met, the associated LCO is not met (i.e., the system is inoperable).

However, the existing TS Bases are contrary to SR 3.0.1 and state that the LCO is met even if the manual actuation SR isn't met. The manual actuation SR is removed to be consistent with the TS requirements.

BWR/4 SR 3.4.3.2 and BWR/6 SR 3.4.4.3 are not needed to verify the S/RVs are mechanically functioning properly. The as-left test of the S/RVs verifies each valve is functioning properly.

Following installation in the plant, the actuator is tested with the actuator disengaged. This test not only verifies operability of the actuator, but also verifies the pneumatic and electric connections. The actuators are then re-engaged to the valve with appropriate independent verification and sign-offs to verify the S/RV will function properly. The STS does not include post-maintenance or post-installation testing, as described in the SR 3.0.1 Bases. Therefore, an SR is not needed for that purpose.

The SR is not needed to verify the downstream piping is unobstructed. Licensees have robust Foreign Material Exclusion (FME) programs to ensure systems are not contaminated during maintenance. Those programs are routinely inspected under NRC Inspection Manual Procedure 71111.20. Further, in licensing actions the staff agreed that an SR to verify that S/RV downstream piping is not obstructed is unnecessary because of licensee FME controls. The safety evaluation for license amendment 116 for the Hope Creek Generating Station, dated February 10, 1999, stated:

Another difference between the current TS-required stroking and the licensee's proposal is that, when performing the testing in-situ as required by the current TS, the testing verifies that the SRV discharge line is not blocked. However, the licensee stated that there is a Foreign Material Exclusion Program in place at the plant which minimizes the potential of debris blocking the discharge lines such that the possibility of blockage is extremely remote. The staff agrees that there is a very small possibility of blockage of an SRV discharge line as demonstrated by operational history and finds that the licensee has acceptably addressed this concern.

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TSTF-576, Rev. 1 As discussed above, there are nine additional BWR unit TS that do not require manual actuation of the safety mode S/RVs. There is no industry operational history of S/RV downstream piping being obstructed by foreign material and the staff's conclusion regarding the Hope Creek FME program is equally applicable to other plants adopting the proposed change.

Testing of S/RV actuation in relief mode is unaffected by the proposed change. BWR/6 SR 3.4.4.2 requires verification that each required S/RV acting in the relief mode actuates on an actual or simulated signal. The initiating instrumentation of the S/RVs in relief mode will continue to be tested by BWR/6 TS 3.3.6.5, "Relief and Low-Low Set (LLS) Instrumentation."

The two BWR/4 plants that credit S/RVs in relief mode will retain similar requirements.

Manual actuation of other S/RV operating modes will continue to be tested by BWR/4 SR 3.5.1.12 and BWR/6 SR 3.5.1.7, "Verify each ADS valve opens when manually actuated,"

and BWR/4 and BWR/6 SR 3.6.1.6.1, "Verify each LLS valve opens when manually actuated."

In conclusion, the SR to perform a manual actuation of the safety mode S/RVs is not necessary to ensure the specified safety function can be performed.

Action Changes Existing BWR/4 TS 3.4.4, Condition A, applies when one or two [required] S/RVs are inoperable. The Action requires the inoperable S/RV(s) to be restored within 14 days, followed by a plant shutdown.

NUREG-1433 and NUREG-1434 contain an Action for one or two inoperable required safety mode S/RVs, but no BWR plant TS contain equivalent requirements. Three plants (Hatch, Hope Creek, and Monticello) contain an Action that permits one or two required S/RVs to be inoperable, but the corresponding LCOs require more S/RVs to be operable than are credited in the overpressure analysis. As a result, for those plants the assumptions of the overpressure analysis can still be met when one or two safety mode S/RVs are inoperable. This is consistent with the STS Required Action A.1 Bases which states, "With the safety function of one [or two]

[required] S/RV[s] inoperable, the remaining OPERABLE S/RVs are capable of providing the necessary overpressure protection. Because of additional design margin, the ASME Code limits for the RCPB can also be satisfied with two S/RVs inoperable. However, the overall reliability of the pressure relief system is reduced because additional failures in the remaining OPERABLE S/RVs could result in failure to adequately relieve pressure during a limiting event. For this reason, continued operation is permitted for a limited time only." (Note that Hatch was the lead plant for the development of NUREG-1433 and this why the Required Action Bases reflect the Hatch requirements.)

The remaining BWR plants' current LCOs only require operability of the number of S/RVs assumed in the overpressure analysis. There are no additional S/RVs required to be operable.

Therefore, if one of the required S/RVs is inoperable, the assumptions of the overpressure analysis cannot be met and a plant shutdown is required. Similar to those plants' TS, the proposed Overpressure Protection System LCO does not require additional S/RVs beyond what is required to perform the overpressure protection function, Required Action A.1 is not applicable, and it is removed.

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TSTF-576, Rev. 1 The existing BWR/6 TS 3.4.4 Actions are not consistent with the plant TS of the four BWR/6 plants. TS 3.4.4 for the BWR/6 plants (River Bend, Grand Gulf, Perry, and Clinton) contains a single action for one or more required safety mode or relief mode S/RVs inoperable, and it requires being in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, which is consistent with the proposed change. Therefore, the BWR/6 STS is revised to be consistent with the plant-specific BWR/6 TS.

Should it be determined that an S/RV is unavailable, an evaluation under the Corrective Action Program would be required to determine whether the Overpressure Protection System is operable. Availability of the S/RVs will also be assessed under 10 CFR 50.65 (the Maintenance Rule).

Existing BWR/4 TS 3.4.4, Condition B, applies when the Required Action and associated Completion Time of Condition A is not met. As Condition A is deleted, Condition B is no longer needed and is also deleted.

Existing BWR/4 TS 3.4.4, Condition C, applies when [three] or more [required] S/RVs are inoperable. The Condition is renumbered Condition A and revised to state, "OPS inoperable."

The existing Required Actions to be in Mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> are retained.

Testing to verify that the OPS is operable is typically performed during a shutdown when the LCO is not applicable. Therefore, it is worthwhile to consider how the Actions would be applied at power. As an example, consider a unit operating at 100% power. A vendor bulletin is received that identifies several installed S/RVs have faulty parts that could reduce the S/RV's relief capacity by 10%. Under the Corrective Action Program, the licensee must evaluate whether the Overpressure Protection System is operable. An evaluation is performed to determine if the reduction in relief capacity in the affected S/RVs would render the OPS incapable of protecting Safety Limit 2.1.2 in the limiting event. If the remaining relief capacity is capable of protecting the Safety Limit, the OPS is operable. If not, the OPS is inoperable, and Action A would require a plant shutdown. Even though the proposed LCO and SRs do not specify the number of required S/RVs or their setpoints, there is no change in the need to ensure the ability to protect Safety Limit 2.1.2 when a condition is identified that could affect the S/RVs. As an additional example, consider a unit operating at 100% power. During the previous outage, all of the S/RVs were removed for testing and replaced with refurbished S/RVs.

The ASME Code testing of the removed S/RVs is completed and the as-found lift settings for some of the S/RVs is not within tolerance. If an evaluation of the deficient condition performed under the Corrective Action Program determines that the collective S/RV performance did not support the previous cycle RLA overpressure analysis, an evaluation will assess the OPS operability for the current cycle. The nature of the evaluation will depend on a number of factors, such as the number of S/RVs that were tested, the number of S/RVs that did not open within tolerance, any known cause of the difference in S/RV performance, the difference between the actual and assumed lift pressure, and any differences between the tested valves and the currently installed valves. The results of the evaluation could result in a spectrum of actions, such as trending and monitoring, revising the RLA overpressure analysis, supplemental S/RV testing, or replacement of the installed valves. If the evaluation determines that the OPS is not operable, the TS Actions require a plant shutdown.

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TSTF-576, Rev. 1

4. REGULATORY EVALUATION 4.1. Applicable Regulatory Requirements/Criteria Section IV, "The Commission Policy," of the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (58 Federal Register 39132), dated July 22, 1993, states in part:

The purpose of Technical Specifications is to impose those conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety by identifying those features that are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval.

[T]he Commission will also entertain requests to adopt portions of the improved STS, even if the licensee does not adopt all STS improvements.

The Commission encourages all licensees who submit Technical Specification related submittals based on this Policy Statement to emphasize human factors principles.

In accordance with this Policy Statement, improved STS have been developed and will be maintained for [BWR designs]. The Commission encourages licensees to use the improved STS as the basis for plant-specific Technical Specifications.

[I]t is the Commission intent that the wording and Bases of the improved STS be used to the extent practicable.

As described in the Commissions "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," recommendations were made by NRC and industry task groups for new STS that include greater emphasis on human factors principles in order to add clarity and understanding to the text of the STS, and provide improvements to the Bases of STS, which provides the purpose for each requirement in the specification. Improved vendor-specific STS were developed and issued by the NRC in September 1992.

The regulation at Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) requires an applicant for an operating license to include in the application proposed TS in accordance with the requirements of 10 CFR 50.36. The applicant must include in the application a "summary statement of the bases or reasons for such specifications, other than those covering administrative controls." However, per 10 CFR 50.36(a)(1), these technical specification bases "shall not become part of the technical specifications." The Final Policy Statement provides the following description of the scope and the purpose of the Technical Specification Bases:

Appropriate Surveillance Requirements and Actions should be retained for each LCO

[limiting condition for operation] which remains or is included in the Technical Specifications. Each LCO, Action, and Surveillance Requirement should have supporting Bases. The Bases should at a minimum address the following questions and Page 16

TSTF-576, Rev. 1 cite references to appropriate licensing documentation (e.g., FSAR, Topical Report) to support the Bases.

1. What is the justification for the Technical Specification, i.e., which Policy Statement criterion requires it to be in the Technical Specifications?

The proposed change does not alter the applicable Policy Statement criterion and justifies that the specification satisfies Criterion 3.

2. What are the Bases for each LCO, i.e., why was it determined to be the lowest functional capability or performance level for the system or component in question necessary for safe operation of the facility and, what are the reasons for the Applicability of the LCO?

The proposed change alters the LCO. The proposed Bases justify why the LCO represents the lowest functional capability or performance level for the system. The proposed change does not alter the Applicability and the Bases continue to justify the Applicability.

3. What are the Bases for each Action, i.e., why should this remedial action be taken if the associated LCO cannot be met; how does this Action relate to other Actions associated with the LCO; and what justifies continued operation of the system or component at the reduced state from the state specified in the LCO for the allowed time period?

The Actions in the proposed change require a plant shutdown if the LCO is not met. The Bases justify that Action and the Completion Time and describe why continued operation in that condition is not permitted.

4. What are the Bases for each Safety Limit?

The proposed change does not alter any Safety Limits or their associated Bases.

5. What are the Bases for each Surveillance Requirement and Surveillance Frequency; i.e., what specific functional requirement is the surveillance designed to verify? Why is this surveillance necessary at the specified frequency to assure that the system or component function is maintained, that facility operation will be within the Safety Limits, and that the LCO will be met?

The proposed change alters the Surveillance Requirements. The proposed associated Bases describe the functional requirement (protecting the overpressure Safety Limit) and why the specified Frequency is appropriate.

Note: In answering these questions the Bases for each number (e.g., Allowable Value, Response Time, Completion Time, Surveillance Frequency), state, condition, and definition (e.g., operability) should be clearly specified. As an example, a number might Page 17

TSTF-576, Rev. 1 be based on engineering judgment, past experience, or PSA [probabilistic safety assessment] insights; but this should be clearly stated.

Additionally, 10 CFR 50.36(b) requires:

Each license authorizing operation of a utilization facility will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to [10 CFR] 50.34 ["Contents of applications; technical information"]. The Commission may include such additional technical specifications as the Commission finds appropriate.

The categories of items required to be in the TS are provided in 10 CFR 50.36(c). As required by 10 CFR 50.36(c)(2)(i), the TS will include LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Per 10 CFR 50.36(c)(2)(i), when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TS until the condition can be met.

The regulation at 10 CFR 50.36(c)(3) requires TS to include items in the category of SRs, which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

Per 10 CFR 50.90, whenever a holder of a license desires to amend the license, application for an amendment must be filed with the Commission, fully describing the changes desired, and following as far as applicable, the form prescribed for original applications.

Per 10 CFR 50.92(a), in determining whether an amendment to a license will be issued to the applicant, the Commission will be guided by the considerations which govern the issuance of initial licenses to the extent applicable and appropriate.

The NRC staffs guidance for the review of TS is in Chapter 16, "Technical Specifications," of NUREG-0800, Revision 3, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (SRP), dated March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared Standard Technical Specifications for each of the light-water reactor nuclear designs.

4.2. Conclusions In conclusion, based on the considerations discussed above, the proposed revision does not alter the current manner of operation and (1) there is reasonable assurance that the health and safety of the public will not be endangered by continued operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

Page 18

TSTF-576, Rev. 1

5. REFERENCES None.

Page 19

TSTF-576, Rev. 0 Model Application

TSTF-576, Rev. 0

[DATE] 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 PLANT NAME DOCKET NO. 50-[xxx]

SUBJECT:

Application to Revise Technical Specifications to Adopt TSTF-576, "Revise Safety/Relief Valve Requirements" Pursuant to 10 CFR 50.90, [LICENSEE] is submitting a request for an amendment to the Technical Specifications (TS) for [PLANT NAME, UNIT NOS.].

[LICENSEE] requests adoption of TSTF-576, "Revise Safety/Relief Valve Requirements." The proposed change revises the Safety/Relief Valve (S/RV) TS to align the overpressure protection requirements with the safety limits and the regulations.

The enclosure provides a description and assessment of the proposed changes. Attachment 1 provides the existing TS pages marked to show the proposed changes. Attachment 2 provides revised (clean) TS pages. Attachment 3 provides the existing TS Bases pages marked to show revised text associated with the proposed TS changes and is provided for information only.

[LICENSEE] requests that the amendment be reviewed under the Consolidated Line Item Improvement Process (CLIIP). Approval of the proposed amendment is requested within six months of acceptance. Once approved, the amendment shall be implemented within [30] days.

There are no regulatory commitments made in this submittal.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated [STATE] Official.

[In accordance with 10 CFR 50.30(b), a license amendment request must be executed in a signed original under oath or affirmation. This can be accomplished by attaching a notarized affidavit confirming the signature authority of the signatory, or by including the following statement in the cover letter: "I declare under penalty of perjury that the foregoing is true and correct.

Executed on (date)." The alternative statement is pursuant to 28 USC 1746. It does not require notarization.]

Page 1

TSTF-576, Rev. 0 If you should have any questions regarding this submittal, please contact [NAME, TELEPHONE NUMBER].

Sincerely,

[Name, Title]

Enclosure:

Description and Assessment Attachments: 1. Proposed Technical Specification Changes (Mark-Up)

2. Revised Technical Specification Pages

3. Proposed Technical Specification Bases Changes (Mark-Up) - For Information Only

[The attachments are to be provided by the licensee and are not included in the model application.]

cc: NRC Project Manager NRC Regional Office NRC Resident Inspector State Contact Page 2

TSTF-576, Rev. 0 ENCLOSURE DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

[LICENSEE] requests adoption of TSTF-576, "Revise Safety/Relief Valve Requirements." The proposed change revises the Safety/Relief Valve (S/RV) Technical Specifications (TS) to align the overpressure protection requirements with the safety limits and the regulations.

2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation

[LICENSEE] has reviewed the safety evaluation for TSTF-576 provided to the Technical Specifications Task Force in a letter dated [DATE]. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-576. As described herein,

[LICENSEE] has concluded that the justifications presented in TSTF-576 and the safety evaluation prepared by the NRC staff are applicable to [PLANT, UNIT NOS.] and justify this amendment for the incorporation of the changes to the [PLANT] TS.

2.2 Optional Changes and Variations

[LICENSEE is not proposing any variations from the TS changes described in TSTF-576 or the applicable parts of the NRC staffs safety evaluation.] [LICENSEE is proposing the following variations from the TS changes described in TSTF-576 or the applicable parts of the NRC staffs safety evaluation: describe the variations]

[The [PLANT] TS utilize different [numbering][and][titles] than the Standard Technical Specifications on which TSTF-576 was based. Specifically, [describe differences between the plant-specific TS numbering and/or titles and the TSTF-576 numbering and titles.] These differences are administrative and do not affect the applicability of TSTF-576 to the [PLANT]

TS.]

[{The following is only applicable to Dresden 2 and 3 and Quad Cities 1 and 2} The [PLANT]

overpressure analysis credits some S/RVs in relief mode in addition to the S/RVs in safety mode to ensure the ASME Code overpressure limit is protected. Existing SR [3.4.4.2] that verifies each required relief function S/RV actuates on an actual or simulated automatic initiation signal is retained. The reference to "relief function" is replaced with "relief mode" to be consistent with the terminology in the Bases.]

[The [PLANT] TS contain requirements that differ from the Standard Technical Specifications on which TSTF-576 was based but are encompassed in the TSTF-576 justification. [Describe differences and why TSTF-576 is still applicable.))

Page 3

TSTF-576, Rev. 0

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis

[LICENSEE] requests adoption of TSTF-576, "Revise Safety/Relief Valve Requirements." The proposed change revises the Safety/Relief Valve (S/RV) Technical Specifications (TS) to align the overpressure protection requirements with the safety limits and the regulations. The Limiting Condition for Operation (LCO) and Surveillance Requirements (SRs) are revised to replace requirements on each credited S/RV with a requirement that the Overpressure Protection System (OPS) be operable. Operability of the OPS is defined as the capability to prevent an overpressure event from exceeding Safety Limit 2.1.2, "Reactor Coolant System Pressure." An SR that tests the ability of the S/RVs to be capable of manual operation is removed as that capability is not credited in any safety analysis. [An SR that verifies the ability of credited S/RVs acting in the relief mode is revised to be consistent with the revised LCO.] The TS Actions are revised to be consistent with the changes to the LCO and SRs. Administrative changes are made to the TS for clarity and consistency.

[LICENSEE] has evaluated if a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change revises the S/RV TS to align the overpressure protection requirements with the safety limits and the regulations. The overpressure protection system must accommodate the most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position). The proposed change does not affect the MSIVs and would have no effect on the probability of the MSIVs closing or generation of a reactor scram signal on MSIV position. Therefore, the probability of the event is unaffected.

The consequences of the accident are based on the peak reactor pressure vessel pressure.

Both the current and proposed TS ensure the overpressure Safety Limit is not exceeded.

The accident analyses consider the aggregate operation of the credited S/RVs, not the performance of individual valves. The proposed change moves the S/RV setpoints and tolerances to licensee control, to be governed by the Inservice Testing Program, which is required by Title 10 of the Code of Federal Regulations, part 50.55a. Altering the control process for these values has no effect on the accident evaluations. As a result, the consequences of the accident are not changed.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Page 4

TSTF-576, Rev. 0

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change revises the S/RV TS to align the overpressure protection requirements with the safety limits and the regulations. The proposed change does not alter the design function or operation of the S/RVs. The proposed change does not create any new credible failure mechanisms, malfunctions, or accident initiators not already considered in the design and licensing basis.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed change revises the S/RV TS to align the overpressure protection requirements with the safety limits and the regulations. The proposed change ensures that the S/RVs can protect Safety Limit 2.1.2. Although the setpoints and tolerances of specific S/RVs are moved to licensee control, the safety margin provided by the aggregate S/RV capability, which ensures the Safety Limit is protected, is not changed.

The conservatisms in the evaluation and the analysis are described in the NRC-approved methods for each licensee, which are not altered by the proposed change. The proposed change does not alter a design basis limit or a safety limit, and, therefore, does not reduce the margin of safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, [LICENSEE] concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

3.2 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

4.0 ENVIRONMENTAL EVALUATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed Page 5

TSTF-576, Rev. 0 amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Page 6

TSTF-576, Rev. 0 Technical Specifications and Bases Changes

TSTF-576, Rev. 1 TABLE OF CONTENTS Page 3.3 INSTRUMENTATION (continued) 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation ......................... 3.3.5.1-1 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation ....................................................................................... 3.3.5.2-1 3.3.5.3 Reactor Core Isolation Cooling (RCIC) System Instrumentation ................. 3.3.5.3-1 3.3.6.1 Primary Containment Isolation Instrumentation ........................................... 3.3.6.1-1 3.3.6.2 Secondary Containment Isolation Instrumentation....................................... 3.3.6.2-1 3.3.6.3 Low-Low Set (LLS) Instrumentation ............................................................. 3.3.6.3-1 3.3.7.1 [ Main Control Room Environmental Control (MCREC) ] System Instrumentation ....................................................................................... 3.3.7.1-1 3.3.8.1 Loss of Power (LOP) Instrumentation .......................................................... 3.3.8.1-1 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring ...................... 3.3.8.2-1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating ........................................................................ 3.4.1-1 3.4.2 Jet Pumps ....................................................................................................... 3.4.2-1 3.4.3 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs) ........ 3.4.3-1 3.4.4 RCS Operational LEAKAGE ........................................................................... 3.4.4-1 3.4.5 RCS Pressure Isolation Valve (PIV) Leakage ................................................. 3.4.5-1 3.4.6 RCS Leakage Detection Instrumentation ........................................................ 3.4.6-1 3.4.7 RCS Specific Activity ....................................................................................... 3.4.7-1 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown .................................................................................................. 3.4.8-1 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown .................................................................................................. 3.4.9-1 3.4.10 RCS Pressure and Temperature (P/T) Limits ............................................... 3.4.10-1 3.4.11 Reactor Steam Dome Pressure .................................................................... 3.4.11-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating ........................................................................................... 3.5.1-1 3.5.2 RPV Water Inventory Control .......................................................................... 3.5.2-1 3.5.3 RCIC System................................................................................................... 3.5.3-1 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary Containment.................................................................................... 3.6.1.1-1 3.6.1.2 Primary Containment Air Lock...................................................................... 3.6.1.2-1 3.6.1.3 Primary Containment Isolation Valves (PCIVs) ............................................ 3.6.1.3-1 3.6.1.4 Drywell Pressure .......................................................................................... 3.6.1.4-1 3.6.1.5 Drywell Air Temperature............................................................................... 3.6.1.5-1 3.6.1.6 Low-Low Set (LLS) Valves ........................................................................... 3.6.1.6-1 General Electric BWR/4 STS vi Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Overpressure Protection System (OPS) Safety/Relief Valves (S/RVs)

LCO 3.4.3 The OPS safety function of [11] S/RVs shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. [ One [or two] [required] A.1 Restore the [required] 14 days S/RV[s] inoperable. S/RV[s] to OPERABLE status. [OR In accordance with the Risk Informed Completion Time Program] ]

B. [ Required Action and B.1 ---------------NOTE--------------

associated Completion LCO 3.0.4.a is not Time of Condition A not applicable when entering met. ] MODE 3.

Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AC. OPS AC.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable[Three] or more [required] S/RVs AND inoperable.

AC.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> General Electric BWR/4 STS 3.4.3-1 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the OPS has the capability to prevent reactor steam dome pressure from exceeding Safety Limit 2.1.2.

---

NOTE------------------------------

[2] [required] S/RVs may be changed to a lower setpoint group. [ In accordance

---

with the INSERVICE Verify the safety function lift setpoints of the TESTING

[required] S/RVs are as follows: PROGRAM Number of Setpoint OR S/RVs (psig)

[ [18] months]

[4] [1090 +/- 32.7]

[4] [1100 +/- 33.0] OR

[3] [1110 +/- 33.3]

In accordance Following testing, lift settings shall be within +/- 1%. with the Surveillance Frequency Control Program ]

SR 3.4.3.2 -------------------------------NOTE------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify each [required] S/RV opens when manually [ [18] months [on actuated. a STAGGERED TEST BASIS for each valve solenoid OR In accordance with the Surveillance Frequency Control Program ] ]

General Electric BWR/4 STS 3.4.3-2 Rev. 5.0

TSTF-576, Rev. 1 TABLE OF CONTENTS Page B 3.3 INSTRUMENTATION (continued)

B 3.3.6.1 Primary Containment Isolation Instrumentation ......................................... B 3.3.6.1-1 B 3.3.6.2 Secondary Containment Isolation Instrumentation..................................... B 3.3.6.2-1 B 3.3.6.3 Low-Low Set (LLS) Instrumentation ........................................................... B 3.3.6.3-1 B 3.3.7.1 [ Main Control Room Environmental Control (MCREC) ]

System Instrumentation.............................................................................. B 3.3.7.1-1 B 3.3.8.1 Loss of Power (LOP) Instrumentation ........................................................ B 3.3.8.1-1 B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring .................... B 3.3.8.2-1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 Recirculation Loops Operating ................................................................... B 3.4.1-1 B 3.4.2 Jet Pumps .................................................................................................. B 3.4.2-1 B 3.4.3 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs) ... B 3.4.3-1 B 3.4.4 RCS Operational LEAKAGE ...................................................................... B 3.4.4-1 B 3.4.5 RCS Pressure Isolation Valve (PIV) Leakage ............................................ B 3.4.5-1 B 3.4.6 RCS Leakage Detection Instrumentation ................................................... B 3.4.6-1 B 3.4.7 RCS Specific Activity .................................................................................. B 3.4.7-1 B 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System

- Hot Shutdown .......................................................................................... B 3.4.8-1 B 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System

- Cold Shutdown ........................................................................................ B 3.4.9-1 B 3.4.10 RCS Pressure and Temperature (P/T) Limits ............................................ B 3.4.10-1 B 3.4.11 Reactor Steam Dome Pressure ................................................................. B 3.4.11-1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS - Operating ...................................................................................... B 3.5.1-1 B 3.5.2 RPV Water Inventory Control ..................................................................... B 3.5.2-1 B 3.5.3 RCIC System.............................................................................................. B 3.5.3-1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment.................................................................................. B 3.6.1.1-1 B 3.6.1.2 Primary Containment Air Lock.................................................................... B 3.6.1.2-1 B 3.6.1.3 Primary Containment Isolation Valves (PCIVs) .......................................... B 3.6.1.3-1 B 3.6.1.4 Drywell Pressure ........................................................................................ B 3.6.1.4-1 B 3.6.1.5 Drywell Air Temperature............................................................................. B 3.6.1.5-1 B 3.6.1.6 Low-Low Set (LLS) Valves ......................................................................... B 3.6.1.6-1 General Electric BWR/4 STS vi Rev. 5.0

TSTF-576, Rev. 1 RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and anticipated operational occurrences (AOOs).

During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core may be done under LCO 3.10.1, "Inservice Leak and Hydrostatic Testing Operation." Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASME Code, Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB, reducing the number of protective barriers designed to prevent radioactive releases from exceeding the limits specified in 10 CFR 100, "Reactor Site Criteria" (Ref. 4). If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere.

APPLICABLE The Overpressure Protection System RCS safety/relief valves and the Reactor Protection System Reactor SAFETY Vessel Steam Dome Pressure - High Function have settings established ANALYSES to ensure that the RCS pressure SL will not be exceeded.

The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME, Boiler and Pressure Vessel Code, [1971 Edition], including Addenda through the [winter of 1972] (Ref. 5), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig.

The SL of 1325 psig, as measured in the reactor steam dome, is General Electric BWR/4 STS B 2.1.2-1 Rev. 5.0

TSTF-576, Rev. 1 Control Rod Scram Times B 3.1.4 BASES APPLICABLE SAFETY ANALYSES (continued) during the control rod drop accident (Ref. 5) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Rod Pattern Control"). For the reactor vessel overpressure protection analysis, the scram function, along with the Overpressure Protection Systemsafety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

Control rod scram times satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The scram times specified in Table 3.1.4-1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met (Ref. 6). To account for single failures and "slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis. The scram times have a margin that allows up to approximately 7% of the control rods (e.g., 137 x 7% = 10) to have scram times exceeding the specified limits (i.e., "slow" control rods) assuming a single stuck control rod (as allowed by LCO 3.1.3, "Control Rod OPERABILITY") and an additional control rod failing to scram per the single failure criterion. The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes ("pickup") when the index tube passes a specific location and then opens ("dropout") as the index tube travels upward. Verification of the specified scram times in Table 3.1.4-1 is accomplished through measurement of the "dropout" times. To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed "slow" control rods may occupy adjacent locations.

Table 3.1.4-1 is modified by two Notes which state that control rods with scram times not within the limits of the Table are considered "slow" and that control rods with scram times > 7 seconds are considered inoperable as required by SR 3.1.3.3.

This LCO applies only to OPERABLE control rods since inoperable control rods will be inserted and disarmed (LCO 3.1.3). Slow scramming control rods may be conservatively declared inoperable and not accounted for as "slow" control rods.

APPLICABILITY In MODES 1 and 2, a scram is assumed to function during transients and accidents analyzed for these plant conditions. These events are assumed to occur during startup and power operation; therefore, the scram function of the control rods is required during these MODES. In MODES 3 and 4, the control rods are not able to be withdrawn since the General Electric BWR/4 STS B 3.1.4-2 Rev. 5.0

TSTF-576, Rev. 1 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) located. Each APRM channel receives two total drive flow signals representative of total core flow. The total drive flow signals are generated by four flow units, two of which supply signals to the trip system A APRMs, while the other two supply signals to the trip system B APRMs. Each flow unit signal is provided by summing up the flow signals from the two recirculation loops. To obtain the most conservative reference signals, the total flow signals from the two flow units (associated with a trip system as described above) are routed to a low auction circuit associated with each APRM. Each APRM's auction circuit selects the lower of the two flow unit signals for use as the scram trip reference for that particular APRM. Each required Average Power Range Monitor Flow Biased Simulated Thermal Power - High channel only requires an input from one OPERABLE flow unit, since the individual APRM channel will perform the intended function with only one OPERABLE flow unit input. However, in order to maintain single failure criteria for the Function, at least one required Average Power Range Monitor Flow Biased Simulated Thermal Power - High channel in each trip system must be capable of maintaining an OPERABLE flow unit signal in the event of a failure of an auction circuit, or a flow unit, in the associated trip system (e.g., if a flow unit is inoperable, one of the two required Average Power Range Monitor Flow Biased Simulated Thermal Power - High channels in the associated trip system must be considered inoperable).

The clamped Allowable Value is based on analyses that take credit for the Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function for the mitigation of the loss of feedwater heating event. The THERMAL POWER time constant of < 7 seconds is based on the fuel heat transfer dynamics and provides a signal proportional to the THERMAL POWER.

The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity.

2.c. Average Power Range Monitor Fixed Neutron Flux - High The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases.

The Average Power Range Monitor Fixed Neutron Flux - High Function is capable of generating a trip signal to prevent fuel damage or excessive General Electric BWR/4 STS B 3.3.1.1-10 Rev. 5.0

TSTF-576, Rev. 1 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

RCS pressure. For the overpressurization protection analysis of Reference 5, the Average Power Range Monitor Fixed Neutron Flux -

High Function is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the Overpressure Protection Systemsafety/relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 6) takes credit for the Average Power Range Monitor Fixed Neutron Flux - High Function to terminate the CRDA.

The APRM System is divided into two groups of channels with three APRM channels inputting to each trip system. The system is designed to allow one channel in each trip system to be bypassed. Any one APRM channel in a trip system can cause the associated trip system to trip.

Four channels of Average Power Range Monitor Fixed Neutron Flux -

High with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. In addition, to provide adequate coverage of the entire core, at least 11 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located.

The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses.

The Average Power Range Monitor Fixed Neutron Flux - High Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Fixed Neutron Flux - High Function is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux - High, Setdown Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor Fixed Neutron Flux - High Function is not required in MODE 2.

2.d. Average Power Range Monitor - Downscale This signal ensures that there is adequate Neutron Monitoring System protection if the reactor mode switch is placed in the run position prior to the APRMs coming on scale. With the reactor mode switch in run, an APRM downscale signal coincident with an associated Intermediate Range Monitor Neutron Flux - High or Inop signal generates a trip signal.

This Function was not specifically credited in the accident analysis but it General Electric BWR/4 STS B 3.3.1.1-11 Rev. 5.0

TSTF-576, Rev. 1 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

3. Reactor Vessel Steam Dome Pressure - High An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure -

High Function initiates a scram for transients that results in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 5, reactor scram (the analyses conservatively assume scram on the Average Power Range Monitor Fixed Neutron Flux - High signal, not the Reactor Vessel Steam Dome Pressure - High signal), along with the Overpressure Protection SystemS/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits.

High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure

- High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.

Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.

4. Reactor Vessel Water Level - Low, Level 3 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level - Low, Level 3 Function is assumed in the analysis of the recirculation line break (Ref. 7). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

General Electric BWR/4 STS B 3.3.1.1-14 Rev. 5.0

TSTF-576, Rev. 1 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Four channels of Reactor Vessel Water Level - Low, Level 3 Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.

The Reactor Vessel Water Level - Low, Level 3 Allowable Value is selected to ensure that during normal operation the separator skirts are not uncovered (this protects available recirculation pump net positive suction head (NPSH) from significant carryunder) and, for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS subsystems at Reactor Vessel Water - Low Low Low, Level 1 will not be required.

The Function is required in MODES 1 and 2 where considerable energy exists in the RCS resulting in the limiting transients and accidents. ECCS initiations at Reactor Vessel Water Level - Low Low, Level 2 and Low Low Low, Level 1 provide sufficient protection for level transients in all other MODES.

5. Main Steam Isolation Valve - Closure MSIV closure results in loss of the main turbine and the condenser as a heat sink for the nuclear steam supply system and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve - Closure signal before the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization transient.

However, for the overpressurization protection analysis of Reference 5, the Average Power Range Monitor Fixed Neutron Flux - High Function, along with the Overpressure Protection SystemS/RVs, limits the peak RPV pressure to less than the ASME Code limits. That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis. Additionally, MSIV closure is assumed in the transients analyzed in Reference 8 (e.g., low steam line pressure, manual closure of MSIVs, high steam line flow). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

MSIV closure signals are initiated from position switches located on each of the eight MSIVs. Each MSIV has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Each inboard and outboard MSIV inputs to a main steam line channel in each trip system, and each of the two trip logics within each RPS trip system receive parallel inputs from two of the four main steam lines. Thus, each General Electric BWR/4 STS B 3.3.1.1-15 Rev. 5.0

TSTF-576, Rev. 1 ATWS-RPT Instrumentation B 3.3.4.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The specific Applicable Safety Analyses and LCO discussions are listed below on a Function by Function basis.

a. Reactor Vessel Water Level - Low Low, Level 2 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the ATWS-RPT System is initiated at Level 2 to aid in maintaining level above the top of the active fuel. The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff.

Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

Four channels of Reactor Vessel Water Level - Low Low, Level 2, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Vessel Water Level - Low Low, Level 2 Allowable Value is chosen so that the system will not be initiated after a Level 3 scram with feedwater still available, and for convenience with the reactor core isolation cooling initiation.

b. Reactor Steam Dome Pressure - High Excessively high RPV pressure may rupture the RCPB. An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Steam Dome Pressure - High Function initiates a RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the Overpressure Protection Systemsafety/relief valves, limits the peak RPV pressure to less than the ASME Section III Code Service Level C limits (1500 psig).

The Reactor Steam Dome Pressure - High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure.

Four channels of Reactor Steam Dome Pressure - High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude General Electric BWR/4 STS B 3.3.4.2-3 Rev. 5.0

TSTF-576, Rev. 1 LLS Instrumentation B 3.3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)

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REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

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]

SR 3.3.6.3.7 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required actuation logic for a specified channel.

The system functional testing performed in LCO 3.4.3, "Overpressure Protection System (OPS) Safety/Relief Valves(S/RVs)" and LCO 3.6.1.8, "Low-Low Set (LLS) Safety/Relief Valves (S/RVs)," for S/RVs overlaps this test to provide complete testing of the assumed safety function.

[ The Frequency of once every 18 months for SR 3.3.6.3.7 is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

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]

REFERENCES 1. FSAR, Figure [ ] .

2. FSAR, Section [5.5.17].

3. GENE-770-06-1, "Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.

General Electric BWR/4 STS B 3.3.6.3-9 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs)

BASES BACKGROUND The Overpressure Protection System (OPS) prevents overpressurization of the nuclear system by discharging reactor steam to the suppression pool. This action protects the reactor coolant pressure boundary (RCPB) from failure which could result in the release of fission products (Ref. 1).

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref. 2) requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, tThe size and number of safety/relief valves (S/RVs) are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).

The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.

The S/RVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the spring loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve. The safety mode is credited for overpressure protection. This satisfies the Code requirement.

In the relief mode (or power actuated mode of operation), a pneumatic piston or cylinder and mechanical linkage assembly are used to open the valve by overcoming the spring force, even with the valve inlet pressure equal to 0 psig. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure reaches the spring lift set pressures.

In the relief mode, valves may be opened manually or automatically at the selected preset pressure. [S/RVs operating in relief mode are not credited for overpressure protection.][Some S/RVs operating in the relief mode are also credited for overpressure protection.]

[Some of the S/RVs operating in relief mode also provide the low-low set relief function, specified in LCO 3.6.1.6, "Low-Low Set (LLS)

Valves," and the Automatic Depressurization System, specified in LCO 3.5.1, "ECCS - Operating." The instrumentation associated with the [relief mode and] low-low set relief function is discussed in the General Electric BWR/4 STS B 3.4.3-1 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.3 Bases for LCO 3.3.6.3, "Low-Low Set (LLS) Instrumentation," and instrumentation for the ADS function is discussed in LCO 3.3.5.1, "Emergency Core Cooling Systems (ECCS) Instrumentation."]

Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool. The S/RVs that provide the relief mode are the low-low set (LLS) valves and the Automatic Depressurization System (ADS) valves. The LLS requirements are specified in LCO 3.6.1.6, "Low-Low Set (LLS) Valves," and the ADS requirements are specified in LCO 3.5.1, "ECCS - Operating."

APPLICABLE The OPS overpressure protection system must accommodate the most SAFETY severe pressurization transient. Evaluations have determined that the ANALYSES most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 1). The S/RV discharge piping is designed to accommodate forces resulting from relief action including interactions with the suppression pool and is supported for reactions due to flow at maximum S/RV discharge capacity so that system integrity is maintained. For the purpose of Tthe overpressure protection analyses, (Ref. 1) assume

[eleven11] S/RVs are assumed to operate in the safety mode of operation [and an additional [seven] S/RVs operate in the relief mode]. The analysis results demonstrate that the OPS design S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig =

1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design design Basis basis Eventevent.

From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above.

Reference 3 2 discusses additional events that are expected to actuate the S/RVs.

The OPS satisfies S/RVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

General Electric BWR/4 STS B 3.4.3-2 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.3 BASES LCO The OPS is OPERABLE when it can ensure that the ASME Code limit on peak reactor pressure, as stated in Safety Limit 2.1.2, will be protected using the safety [and relief] mode[s] function of the

[11] S/RVs [and the relief mode of additional S/RVs]. are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. 1 and 2). The requirements of this LCO are applicable only toThe OPERABILITY of the OPS is only dependent on the ability capability of the S/RVs to mechanically open to relieve excess pressure and maintain reactor pressure below Safety Limit 2.1.2, and may credit less than the full complement of installed S/RVs. when the lift setpoint is exceeded (safety function).

The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve to be set setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressureization conditions. The transient evaluations in Reference 3 the FSAR are based on these setpoints, but also include the additional uncertainties of +/- 1% of the nominal setpoint drift to provide an added degree of conservatism.

An inoperable OPS Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in Safety Limit 2.1.2 the ASME Code limit on reactor pressure being exceeded.

APPLICABILITY In MODES 1, 2, and 3, the OPS all S/RVs must be OPERABLE, since there may be considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES.

The OPS S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.

In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents. In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The OPS S/RV function is not needed during these conditions.

ACTIONS [ A.1 With the safety function of one [or two] [required] S/RV[s] inoperable, the remaining OPERABLE S/RVs are capable of providing the necessary overpressure protection. Because of additional design margin, the ASME Code limits for the RCPB can also be satisfied with two S/RVs inoperable.

General Electric BWR/4 STS B 3.4.3-3 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.3 However, the overall reliability of the pressure relief system is reduced because additional failures in the remaining OPERABLE S/RVs could result in failure to adequately relieve pressure during a limiting event. For this reason, continued operation is permitted for a limited time only.

BASES ACTIONS (continued)

The 14 day Completion Time to restore the inoperable required S/RVs to OPERABLE status is based on the relief capability of the remaining S/RVs, the low probability of an event requiring S/RV actuation, and a reasonable time to complete the Required Action. [Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.] ]

B.1

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REVIEWERS NOTE ----------------------------------

Adoption of a MODE 3 end state requires the licensee to make the following commitments:

1. [LICENSEE] will follow the guidance established in Section 11 of NUMARC 93-01, "Industry Guidance for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Nuclear Management and Resource Council, Revision [4F].

2. [LICENSEE] will follow the guidance established in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 2, "Technical Specifications End States, NEDC-32988-A," November 2009.

If the safety function of the inoperable required S/RVs cannot be restored to OPERABLE status within the associated Completion Time of Required Action A.1, the plant must be brought to a MODE in which overall plant risk is minimized. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 3) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low risk state.

Required Action B.1 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 3. This Note prohibits the use of LCO 3.0.4.a to enter MODE 3 during startup with the LCO not met. However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 3, and establishment of risk management actions, if appropriate. LCO 3.0.4 is General Electric BWR/4 STS B 3.4.3-4 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.3 BASES ACTIONS (continued) not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

The allowed Completion Time is reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

AC.1 and AC.2 If the OPS is inoperable, [three] or more [required] S/RVs are inoperable, a transient may result in the violation of the ASME Code limit on reactor pressure. The plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillance verifies that the OPS has the capability to prevent the reactor steam dome pressure from exceeding Safety Limit 2.1.2requires that the [required] S/RVs will open at the pressures assumed in the safety analysis of Reference 1. The testing of the demonstration of the S/RV safety mode safe lift settings is must be performed during shutdown, since this is a bench test, [to be done in accordance with the INSERVICE TESTING PROGRAM]. The measured S/RV mechanical lift pressures determined in accordance with the INSERVICE TESTING PROGRAM are reviewed and compared to the overpressure analysis to verify that the collective performance of the credited S/RVs will ensure Safety Limit 2.1.2 is protected.

Should one or more of the credited S/RVs not actuate within the assumed tolerance during as-found testing, the actual lift values will be used to evaluate the affected previous cycle overpressure analyses to determine whether the Safety Limit was protected. In this case, the SR is met by a combination of testing and calculation.

The effect on OPS operability of S/RVs not actuating within the assumed tolerance during as-found testing will be evaluated for the current cycle under the corrective action program using the NRC-approved methods for overpressure and accident analyses.

The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is +/- [3]% for OPERABILITY; however, the valves are reset to General Electric BWR/4 STS B 3.4.3-5 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.3

+/- 1% during the Surveillance to allow for drift. [A Note is provided to allow up to [two] of the required [11] S/RVs to be physically replaced with S/RVs with lower setpoints. This provides operational flexibility which maintains the assumptions in the over-pressure analysis.]

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REVIEWERS NOTE-----------------------------------

If the testing is within the scope of the licensee's INSERVICE TESTING PROGRAM, the Frequency "In accordance with the INSERVICE TESTING PROGRAM" should be used. Otherwise, the periodic Frequency of 18 months or the reference to the Surveillance Frequency Control Program should be used.

[ The 18 month Frequency was selected because this Surveillance must be performed during shutdown conditions and is based on the time between refuelings.

BASES SURVEILLANCE REQUIREMENTS (continued)

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

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]

SR 3.4.3.2 A manual actuation of each [required] S/RV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow.

Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the S/RVs divert steam flow upon opening.

Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at which this test is to be performed is [920] psig (the pressure recommended by the valve manufacturer). Adequate steam flow is represented by [at least 1.25 turbine bypass valves open, or total steam flow 106 lb/hr]. Plant startup is allowed prior to performing this test because valve OPERABILITY and General Electric BWR/4 STS B 3.4.3-6 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.3 the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the S/RV is considered OPERABLE.

[ The [18] month on a STAGGERED TEST BASIS Frequency ensures that each solenoid for each S/RV is alternately tested. The 18 month Frequency was developed based on the S/RV tests required by the ASME Boiler and Pressure Vessel Code (Ref. 4). Operating experience has shown that these components usually pass the Surveillance when General Electric BWR/4 STS B 3.4.3-7 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.3 BASES SURVEILLANCE REQUIREMENTS (continued) performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

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]

REFERENCES 1. FSAR, Section [5.2.2.2.4].

24. ASME Code for Operation and Maintenance of Nuclear Power Plants.

32. FSAR, Section [15].

3. NEDC-32988-A, Revision 2, Technical Justification to Support Risk-Informed Modification to Selected Required End States for BWR Plants, December 2002.

4. ASME Code for Operation and Maintenance of Nuclear Power Plants.

General Electric BWR/4 STS B 3.4.3-8 Rev. 5.0

TSTF-576, Rev. 1 Reactor Steam Dome Pressure B 3.4.11 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.11 Reactor Steam Dome Pressure BASES BACKGROUND The reactor steam dome pressure is an assumed initial condition of design basis accidents and transients and is also an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria.

APPLICABLE The reactor steam dome pressure of [1020] psig is an initial condition of SAFETY the vessel overpressure protection analysis of Reference 1. This ANALYSES analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the Overpressure Protection System pressure relief system, primarily the safety/relief valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved.

Reference 2 also assumes an initial reactor steam dome pressure for the analysis of design basis accidents and transients used to determine the limits for fuel cladding integrity (see Bases for LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)").

Reactor steam dome pressure satisfies the requirements of Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The specified reactor steam dome pressure limit of [1020] psig ensures the plant is operated within the assumptions of the transient analyses.

Operation above the limit may result in a transient response more severe than analyzed.

APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these MODES, the reactor may be generating significant steam and the design basis accidents and transients are bounding.

In MODES 3, 4, and 5, the limit is not applicable because the reactor is shut down. In these MODES, the reactor pressure is well below the required limit, and no anticipated events will challenge the overpressure limits.

General Electric BWR/4 STS B 3.4.11-1 Rev. 5.0

TSTF-576, Rev. 1 LLS Valves B 3.6.1.6 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.6 Low-Low Set (LLS) Valves BASES BACKGROUND The safety/relief valves (S/RVs) can actuate in either the safety mode as part of the Overpressure Protection System, the Automatic Depressurization System mode, or the LLS mode. In the LLS mode (or power actuated mode of operation), a pneumatic diaphragm and stem assembly overcomes the spring force and opens the pilot valve. As in the safety mode, opening the pilot valve allows a differential pressure to develop across the main valve piston and opens the main valve. The main valve can stay open with valve inlet steam pressure as low as

[50] psig. Below this pressure, steam pressure may not be sufficient to hold the main valve open against the spring force of the pilot valves. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure exceeds the safety mode pressure setpoints.

[Four] of the S/RVs are equipped to provide the LLS function. The LLS logic causes the LLS valves to be opened at a lower pressure than the relief or safety mode pressure setpoints and stay open longer, so that reopening more than one S/RV is prevented on subsequent actuations.

Therefore, the LLS function prevents excessive short duration S/RV cycles with valve actuation at the relief setpoint.

Each S/RV discharges steam through a discharge line and quencher to a location near the bottom of the suppression pool, which causes a load on the suppression pool wall. Actuation at lower reactor pressure results in a lower load.

APPLICABLE The LLS relief mode functions to ensure that the containment design SAFETY basis of one S/RV operating on "subsequent actuations" is met. In other ANALYSES words, multiple simultaneous openings of S/RVs (following the initial opening), and the corresponding higher loads, are avoided. The safety analysis demonstrates that the LLS functions to avoid the induced thrust loads on the S/RV discharge line resulting from "subsequent actuations" of the S/RV during Design Basis Accidents (DBAs). Furthermore, the LLS function justifies the primary containment analysis assumption that simultaneous S/RV openings occur only on the initial actuation for DBAs.

Even though [four] LLS S/RVs are specified, all [four] LLS S/RVs do not operate in any DBA analysis.

LLS valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

General Electric BWR/4 STS B 3.6.1.6-1 Rev. 5.0

TSTF-576, Rev. 1 TABLE OF CONTENTS Page 3.3 INSTRUMENTATION (continued) 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation ......................... 3.3.5.1-1 3.3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation ....................................................................................... 3.3.5.2-1 3.3.5.3 Reactor Core Isolation Cooling (RCIC) System Instrumentation ................. 3.3.5.3-1 3.3.6.1 Primary Containment Isolation Instrumentation ........................................... 3.3.6.1-1 3.3.6.2 Secondary Containment Isolation Instrumentation....................................... 3.3.6.2-1 3.3.6.3 Residual Heat Removal (RHR) Containment Spray System Instrumentation ....................................................................................... 3.3.6.3-1 3.3.6.4 Suppression Pool Makeup (SPMU) System Instrumentation ....................... 3.3.6.4-1 3.3.6.5 Relief and Low-Low Set (LLS) Instrumentation ............................................ 3.3.6.5-1 3.3.7.1 [Control Room Fresh Air (CRFA)] System Instrumentation .......................... 3.3.7.1-1 3.3.8.1 Loss of Power (LOP) Instrumentation .......................................................... 3.3.8.1-1 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring ...................... 3.3.8.2-1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating ........................................................................ 3.4.1-1 3.4.2 Flow Control Valves (FCVs) ............................................................................ 3.4.2-1 3.4.3 Jet Pumps ....................................................................................................... 3.4.3-1 3.4.4 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs) ........ 3.4.4-1 3.4.5 RCS Operational LEAKAGE ........................................................................... 3.4.5-1 3.4.6 RCS Pressure Isolation Valve (PIV) Leakage ................................................. 3.4.6-1 3.4.7 RCS Leakage Detection Instrumentation ........................................................ 3.4.7-1 3.4.8 RCS Specific Activity ....................................................................................... 3.4.8-1 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown .................................................................................................. 3.4.9-1 3.4.10 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown ................................................................................................ 3.4.10-1 3.4.11 RCS Pressure and Temperature (P/T) Limits ............................................... 3.4.11-1 3.4.12 Reactor Steam Dome Pressure .................................................................... 3.4.12-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating ........................................................................................... 3.5.1-1 3.5.2 RPV Water Inventory Control ......................................................................... 3.5.2-1 3.5.3 RCIC System................................................................................................... 3.5.3-1 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary Containment.................................................................................... 3.6.1.1-1 3.6.1.2 Primary Containment Air Locks .................................................................... 3.6.1.2-1 3.6.1.3 Primary Containment Isolation Valves (PCIVs) ............................................ 3.6.1.3-1 General Electric BWR/6 STS vi Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 Overpressure Protection System (OPS) Safety/Relief Valves (S/RVs)

LCO 3.4.4 The OPS The safety function of [seven] S/RVs shall be OPERABLE.,

AND The relief function of [seven] additional S/RVs shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. [ One [required] S/RV A.1 Restore [required] S/RV to 14 days inoperable. OPERABLE status.

[OR In accordance with the Risk Informed Completion Time Program] ]

B. [ Required Action and B.1 --------------NOTE--------------

associated Completion LCO 3.0.4.a is not Time of Condition A not applicable when entering met. ] MODE 3.

Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AC. OPS AC.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable[Two] or more [required] S/RVs AND inoperable.

AC.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> General Electric BWR/6 STS 3.4.4-1 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs 3.4.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify the OPS has the capability to prevent reactor steam dome pressure from exceeding Safety Limit 2.1.2.

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NOTE------------------------------

[2] [required] S/RVs may be changed to a lower setpoint group. [ In accordance

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with the INSERVICE Verify the safety function lift setpoints of the TESTING

[required] S/RVs are as follows: PROGRAM Number of Setpoint OR S/RVs (psig)

((18] months]

[8] [1165 +/- 34.9]

[6] [1180 +/- 35.4] OR

[6] [1190 +/- 35.7]

In accordance Following testing, lift settings shall be within +/- 1%. with the Surveillance Frequency Control Program ]

SR 3.4.4.2 -------------------------------NOTE------------------------------

Valve actuation may be excluded.

Verify each [required] safety/relief valve acting in [ [18] months the relief mode relief function S/RV actuates on an actual or simulated automatic initiation signal. OR In accordance with the Surveillance Frequency Control Program ]

General Electric BWR/6 STS 3.4.4-2 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs 3.4.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.4.3 -------------------------------NOTE------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify each [required] S/RV opens when manually [ [18] months on a actuated. STAGGERED TEST BASIS for each valve solenoid OR In accordance with the Surveillance Frequency Control Program ]

General Electric BWR/6 STS 3.4.4-3 Rev. 5.0

TSTF-576, Rev. 1 TABLE OF CONTENTS Page B 3.3 INSTRUMENTATION (continued)

B 3.3.6.1 Primary Containment Isolation Instrumentation ......................................... B 3.3.6.1-1 B 3.3.6.2 Secondary Containment Isolation Instrumentation..................................... B 3.3.6.2-1 B 3.3.6.3 Residual Heat Removal (RHR) Containment Spray System Instrumentation........................................................................................... B 3.3.6.3-1 B 3.3.6.4 Suppression Pool Makeup (SPMU) System Instrumentation ..................... B 3.3.6.4-1 B 3.3.6.5 Relief and Low-Low Set (LLS) Instrumentation .......................................... B 3.3.6.5-1 B 3.3.7.1 [Control Room Fresh Air (CRFA)] System Instrumentation ........................ B 3.3.7.1-1 B 3.3.8.1 Loss of Power (LOP) Instrumentation ........................................................ B 3.3.8.1-1 B 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring .................... B 3.3.8.2-1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 Recirculation Loops Operating ................................................................... B 3.4.1-1 B 3.4.2 Flow Control Valves (FCVs) ....................................................................... B 3.4.2-1 B 3.4.3 Jet Pumps .................................................................................................. B 3.4.3-1 B 3.4.4 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs) ... B 3.4.4-1 B 3.4.5 RCS Operational LEAKAGE ...................................................................... B 3.4.5-1 B 3.4.6 RCS Pressure Isolation Valve (PIV) Leakage ............................................ B 3.4.6-1 B 3.4.7 RCS Leakage Detection Instrumentation ................................................... B 3.4.7-1 B 3.4.8 RCS Specific Activity .................................................................................. B 3.4.8-1 B 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System

- Hot Shutdown ......................................................................................... B 3.4.9-1 B 3.4.10 Residual Heat Removal (RHR) Shutdown Cooling System

- Cold Shutdown ....................................................................................... B 3.4.10-1 B 3.4.11 RCS Pressure and Temperature (P/T) Limits ............................................ B 3.4.11-1 B 3.4.12 Reactor Steam Dome Pressure ................................................................. B 3.4.12-1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS - Operating ...................................................................................... B 3.5.1-1 B 3.5.2 RPV Water Inventory Control ..................................................................... B 3.5.2-1 B 3.5.3 RCIC System.............................................................................................. B 3.5.3-1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment.................................................................................. B 3.6.1.1-1 B 3.6.1.2 Primary Containment Air Locks .................................................................. B 3.6.1.2-1 B 3.6.1.3 Primary Containment Isolation Valves (PCIVs) .......................................... B 3.6.1.3-1 B 3.6.1.4 Primary Containment Pressure .................................................................. B 3.6.1.4-1 B 3.6.1.5 Primary Containment Air Temperature....................................................... B 3.6.1.5-1 B 3.6.1.6 Low-Low Set (LLS) Valves ......................................................................... B 3.6.1.6-1 B 3.6.1.7 Residual Heat Removal (RHR) Containment Spray System ...................... B 3.6.1.7-1 General Electric BWR/6 STS vi Rev. 5.0

TSTF-576, Rev. 1 RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. According to 10 CFR 50, Appendix A, GDC 14, "Reactor Coolant Pressure Boundary," and GDC 15, "Reactor Coolant System Design" (Ref. 1), the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and anticipated operational occurrences (AOOs).

During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core may be done under LCO 3.10.1, "Inservice Leak and Hydrostatic Testing Operation." Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASME Code, Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB, reducing the number of protective barriers designed to prevent radioactive releases from exceeding the limits specified in 10 CFR 100, "Reactor Site Criteria" (Ref. 4). If this occurred in conjunction with a fuel cladding failure, the number of protective barriers designed to prevent radioactive releases from exceeding the limits would be reduced.

APPLICABLE The Overpressure Protection System RCS safety/relief valves and the Reactor Protection System Reactor SAFETY Vessel Steam Dome Pressure - High Function have settings established ANALYSES to ensure that the RCS pressure SL will not be exceeded.

The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to ASME, Boiler and Pressure Vessel Code, Section III, [1971 Edition], including Addenda through the [winter of 1972] (Ref. 5), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig. The SL of General Electric BWR/6 STS B 2.1.2-1 Rev. 5.0

TSTF-576, Rev. 1 Control Rod Scram Times B 3.1.4 BASES APPLICABLE SAFETY ANALYSES (continued) the scram function is assumed to perform during the control rod drop accident (Ref. 6) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Rod Pattern Control"). For the reactor vessel overpressure protection analysis, the scram function, along with the Overpressure Protection Systemsafety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

Control rod scram times satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The scram times specified in Table 3.1.4-1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met. To account for single failure and "slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis. The scram times have a margin to allow up to 7.5% of the control rods (e.g., 193 x 7.5%

= 14) to have scram times that exceed the specified limits (i.e., "slow" control rods) assuming a single stuck control rod (as allowed by LCO 3.1.3, "Control Rod OPERABILITY") and an additional control rod failing to scram per the single failure criterion. The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes ("pickup")

when the index tube passes a specific location and then opens

("dropout") as the index tube travels upward. Verification of the specified scram times in Table 3.1.4-1 is accomplished through measurement of the "dropout" times.

To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed "slow" control rods may occupy adjacent locations.

Table 3.1.4-1 is modified by two Notes, which state control rods with scram times not within the limits of the Table are considered "slow" and that control rods with scram times > [ ] seconds are considered inoperable as required by SR 3.1.3.3.

This LCO applies only to OPERABLE control rods since inoperable control rods will be inserted and disarmed (LCO 3.1.3). Slow scramming control rods may be conservatively declared inoperable and not accounted for as "slow" control rods.

General Electric BWR/6 STS B 3.1.4-2 Rev. 5.0

TSTF-576, Rev. 1 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) trip. Six channels of Average Power Range Monitor Flow Biased Simulated Thermal Power - High, with three channels in each trip system arranged in one-out-of-three logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. In addition, to provide adequate coverage of the entire core, at least 11 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axial levels at which the LPRMs are located. Each APRM channel receives one total drive flow signal representative of total core flow. The recirculation loop drive flow signals are generated by eight flow units. One flow unit from each recirculation loop is provided to each APRM channel. Total drive flow is determined by each APRM by summing up the flow signals provided to the APRM from the two recirculation loops.

The clamped Allowable Value is based on analyses that take credit for the Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function for the mitigation of the loss of feedwater heater event. The THERMAL POWER time constant of < 7 seconds is based on the fuel heat transfer dynamics and provides a signal that is proportional to the THERMAL POWER.

The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity.

2.c. Average Power Range Monitor Fixed Neutron Flux - High The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases.

The Average Power Range Monitor Fixed Neutron Flux - High Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 3, the Average Power Range Monitor Fixed Neutron Flux -

High Function is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the Overpressure Protection Systemsafety/relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 8) takes credit for the Average Power Range Monitor Fixed Neutron Flux - High Function to terminate the CRDA.

General Electric BWR/6 STS B 3.3.1.1-10 Rev. 5.0

TSTF-576, Rev. 1 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

3. Reactor Vessel Steam Dome Pressure - High An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure -

High Function initiates a scram for transients that results in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 3, the reactor scram (the analyses conservatively assume scram on the Average Power Range Monitor Fixed Neutron Flux - High signal, not the Reactor Vessel Steam Dome Pressure-High signal), along with the Overpressure Protection SystemS/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits.

High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure

- High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.

Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.

4. Reactor Vessel Water Level - Low, Level 3 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level - Low, Level 3 Function is assumed in the analysis of the recirculation line break (Ref. 4). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

General Electric BWR/6 STS B 3.3.1.1-13 Rev. 5.0

TSTF-576, Rev. 1 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

6. Main Steam Isolation Valve - Closure MSIV closure results in loss of the main turbine and the condenser as a heat sink for the Nuclear Steam Supply System and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve - Closure signal before the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization transient.

However, for the overpressurization protection analysis of Reference 3, the Average Power Range Monitor Fixed Neutron Flux - High Function, along with the Overpressure Protection SystemS/RVs, limits the peak RPV pressure to less than the ASME Code limits. That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis.

Additionally, MSIV closure is assumed in the transients analyzed in Reference 5 (e.g., low steam line pressure, manual closure of MSIVs, high steam line flow). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

MSIV closure signals are initiated from position switches located on each of the eight MSIVs. Each MSIV has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Each inboard and outboard MSIV inputs to a main steam line channel in each trip system, and each of the two trip logics within each RPS trip system receive parallel inputs from two of the four main steam lines. Thus, each RPS trip system receives an input from eight Main Steam Isolation Valve

- Closure channels, each consisting of one position switch. The logic for the Main Steam Isolation Valve - Closure Function is arranged such that either the inboard or outboard valve on both of the main steam lines (MSLs) in one of the two logics in each RPS trip system must close in order for a scram to occur.

The Main Steam Isolation Valve - Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient.

Sixteen channels of the Main Steam Isolation Valve - Closure Function with eight channels in each trip system are required to be OPERABLE to ensure that no single instrument failure will preclude the scram from this Function on a valid signal. This Function is only required in MODE 1 since, with the MSIVs open and the heat generation rate high, a pressurization transient can occur if the MSIVs close. In MODE 2, the General Electric BWR/6 STS B 3.3.1.1-15 Rev. 5.0

TSTF-576, Rev. 1 ATWS-RPT Instrumentation B 3.3.4.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The specific Applicable Safety Analyses and LCO discussions are listed below on a Function by Function basis.

a. Reactor Vessel Water Level - Low Low, Level 2 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the ATWS-RPT System is initiated at Level 2 to aid in maintaining level above the top of the active fuel. The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff.

Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

Four channels of Reactor Vessel Level - Low Low, Level 2, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Vessel Water Level - Low Low, Level 2, Allowable Value is chosen so that the system will not initiate after a Level 3 scram with feedwater still available, and for convenience with the reactor core isolation cooling (RCIC) initiation.

b. Reactor Steam Dome Pressure - High Excessively high RPV pressure may rupture the RCPB. An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and RPV overpressurization. The Reactor Steam Dome Pressure - High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the Overpressure Protection Systemsafety/relief valves (S/RVs), limits the peak RPV pressure to less than the ASME Section III Code Service Level C limits (1500 psig).

The Reactor Steam Dome Pressure - High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure. Four channels of Reactor Steam Dome Pressure - High, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this General Electric BWR/6 STS B 3.3.4.2-3 Rev. 5.0

TSTF-576, Rev. 1 Relief and LLS Instrumentation B 3.3.6.5 B 3.3 INSTRUMENTATION B 3.3.6.5 Relief and Low-Low Set (LLS) Instrumentation BASES BACKGROUND The Overpressure Protection System safety/relief valves (S/RVs) prevents overpressurization of the nuclear steam system utilizing the safety/relief valves (S/RVs). Instrumentation is provided to support two modes of S/RV operation - the relief function (all valves) and the LLS function (selected valves). Refer to LCO 3.4.4, "Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs)," and LCO 3.6.1.6, "Low-Low Set (LLS) Safety/Relief Valves (S/RVs)," for Applicability Bases for additional information of these modes of S/RV operation. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the S/RV Safety/Relief valve instrumentation, as well as LCOs on other reactor system parameters, and equipment performance.

Technical Specifications are required by 10 CFR 50.36 to include LSSS for variables that have significant safety functions. LSSS are defined by the regulation as "Where a LSSS is specified for a variable on which a safety limit has been placed, the setting must be chosen so that automatic protective actions will correct the abnormal situation before a Safety Limit (SL) is exceeded." The Analytical Limit is the limit of the process variable at which a safety action is initiated, as established by the safety analysis, to ensure that an SL is not exceeded. Any automatic protection action that occurs on reaching the Analytical Limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protection channels must be chosen to be more conservative than the Analytical Limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur.

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REVIEWER'S NOTE-----------------------------------

The term "Limiting Trip Setpoint" [LTSP] is generic terminology for the calculated trip setting (setpoint) value calculated by means of the plant specific setpoint methodology documented in a document controlled under 10 CFR 50.59. The term [LTSP] indicates that no additional margin has been added between the Analytical Limit and the calculated trip setting.

"Nominal Trip Setpoint [NTSP]" is the suggested terminology for the actual setpoint implemented in the plant surveillance procedures where margin has been added to the calculated [LTSP]. The as-found and as-left tolerances will apply to the [NTSP] implemented in the Surveillance procedures to confirm channel performance.

General Electric BWR/6 STS B 3.3.6.5-1 Rev. 5.0

TSTF-576, Rev. 1 Relief and LLS Instrumentation B 3.3.6.5 BASES BACKGROUND (continued) has been found to be different from the [LTSP] due to some drift of the setting may still be OPERABLE because drift is to be expected. This expected drift would have been specifically accounted for in the setpoint methodology for calculating the [LTSP] and thus the automatic protective action would still have ensured that the SL would not be exceeded with the "as found" setting of the protection channel. Therefore, the channel would still be OPERABLE because it would have performed its safety function and the only corrective action required would be to reset the channel within the established as-left tolerance around the [LTSP] to account for further drift during the next surveillance interval. Note that, although the channel is OPERABLE under these circumstances, the trip setpoint must be left adjusted to a value within the as-left tolerance, in accordance with uncertainty assumptions stated in the referenced setpoint methodology (as-left criteria), and confirmed to be operating within the statistical allowances of the uncertainty terms assigned (as-found criteria).

However, there is also some point beyond which the channel may not be able to perform its function due to, for example, greater than expected drift. This value needs to be specified in the Technical Specifications in order to define OPERABILITY of the channels and is designated as the Allowable Value.

If the actual setting (as-found setpoint) of the channel is found to be conservative with respect to the Allowable Value but is beyond the as-found tolerance band, the channel is OPERABLE, but degraded. The degraded condition will be further evaluated during performance of the SR. This evaluation will consist of resetting the channel setpoint to the

[Nominal Trip Setpoint (NTSP)] (within the allowed tolerance), and evaluating the channel response. If the channel is functioning as required and is expected to pass the next surveillance, then the channel is OPERABLE and can be restored to service at the completion of the surveillance. After the surveillance is completed, the channel as-found condition will be entered into the Corrective Action Program for further evaluation.

The Overpressure Protection System, which utilizes the relief function of the S/RVs, prevents overpressurization of the nuclear steam system.

The LLS function of the S/RVs is designed to mitigate the effects of postulated thrust loads on the S/RV discharge lines by preventing subsequent actuations with an elevated water leg in the S/RV discharge line. It also mitigates the effects of postulated pressure loads on the containment by preventing multiple actuations in rapid succession of the S/RVs subsequent to their initial actuation.

General Electric BWR/6 STS B 3.3.6.5-3 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.4 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.4 Overpressure Protection System (OPS)Safety/Relief Valves (S/RVs)

BASES BACKGROUND The Overpressure Protection System (OPS) prevents overpressurization of the nuclear system by discharging reactor steam to the suppression pool. This action protects the reactor coolant pressure boundary (RCPB) from failure which could result in the release of fission products (Ref. 1).

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref. 21) requires the rReactor pPressure vVessel be protected from overpressure during upset conditions by self- actuated safety valves. As part of the nuclear pressure relief system, tThe size and number of safety/relief valves (S/RVs) are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).

The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.

The S/RVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the direct action of the steam pressure in the main steam lines will act against a spring loaded disk that will pop open when the valve inlet pressure exceeds the spring force. The safety mode is credited for overpressure protection.

In the relief mode (or power actuated mode of operation), a pneumatic piston or cylinder and mechanical linkage assembly are used to open the valve by overcoming the spring force, even with the valve inlet pressure equal to 0 psig. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure reaches the spring lift set pressures. In the relief mode, valves may be opened manually or automatically at the selected preset pressure. Some S/RVs operating in the relief mode are also credited for overpressure protection.

Some Six of the S/RVs operating in relief mode providing the relief function also provide the low-low set relief function specified in LCO 3.6.1.6, "Low-Low Set (LLS) Valves,." and Eight of the S/RVs that provide the relief function are part of the Automatic Depressurization System specified in LCO 3.5.1, "ECCS - Operating." The instrumentation associated with the relief valve mode function and low-low set relief function is discussed in the Bases for LCO 3.3.6.5, "Relief and Low-Low General Electric BWR/6 STS B 3.4.4-1 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.4 Set (LLS) Instrumentation," and instrumentation for the ADS function is discussed in LCO 3.3.5.1, "Emergency Core Cooling Systems (ECCS)

Instrumentation."

APPLICABLE The OPS overpressure protection system must accommodate the most SAFETY severe pressurizatione transient. Evaluations have determined that the most ANALYSES most severe transient is the closure of all main steam isolation valves (MSIVs) followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 12). The S/RV discharge piping is designed to accommodate forces resulting from relief action including interactions with the suppression pool and is supported for reactions due to flow at maximum S/RV discharge capacity so that system integrity is maintained. For the purpose of Tthe overpressure protection analyses (Ref. 1), assume [seven]

S/RVs operate in the safety mode of operation and an additional

[seven] S/RVs operate in the relief mode. , [six] of the S/RVs are assumed to operate in the relief mode, and seven in the safety mode.

The analysis results demonstrate that the OPS design S/RV capacity is capable of maintaining reactor pressure below BASES APPLICABLE SAFETY ANALYSES (continued) the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig = 1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the design basis event.

From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above. Reference 3 discusses additional events that are expected to actuate the S/RVs.

The OPS satisfies S/RVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The OPS is OPERABLE when it can ensure that the ASME Code limit on peak reactor pressure, as stated in Safety Limit 2.1.2, will be protected using the safety and relief modes function of the seven S/RVs and the relief mode of additional S/RVs. is required to be OPERABLE in the safety mode, and an additional seven S/RVs (other than the seven S/RVs that satisfy the safety function) must be OPERABLE in the relief mode. The OPERABILITY of the OPS is only dependent on The requirements of this LCO are applicable only to the capability ability of the S/RVs to mechanically open to relieve excess pressure and maintain reactor pressure below Safety Limit 2.1.2, and may credit less than the full complement of installed S/RVs. In Reference 2, an evaluation was performed to establish the parametric relationship between the peak vessel pressure and the number of OPERABLE S/RVs. The results show that with a minimum of seven General Electric BWR/6 STS B 3.4.4-2 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.4 S/RVs in the safety mode and six S/RVs in the relief mode OPERABLE, the ASME Code limit of 1375 psig is not exceeded.

The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve to be set at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressure conditions. The transient evaluations in Reference 3 are based on these setpoints, but also include the additional uncertainties of +/- 1% of the nominal setpoint to account for potential setpoint drift to provide an added degree of conservatism.

An inoperable OPS Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in Safety Limit 2.1.2 the ASME Code limit on reactor pressure being exceeded.

APPLICABILITY In MODES 1, 2, and 3, the OPS specified number of S/RVs must be OPERABLE since there may be considerable energy in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The OPS S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.

BASES APPLICABILITY (continued)

In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents. In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The OPS S/RV function is not needed during these conditions.

ACTIONS A.1 With the safety function of one [required] S/RV inoperable, the remaining OPERABLE S/RVs are capable of providing the necessary overpressure protection. Because of additional design margin, the ASME Code limits for the RCPB can also be satisfied with two S/RVs inoperable. However, the overall reliability of the pressure relief system is reduced because additional failures in the remaining OPERABLE S/RVs could result in failure to adequately relieve pressure during a limiting event. For this reason, continued operation is permitted for a limited time only.

The 14 day Completion Time to restore the inoperable required S/RVs to OPERABLE status is based on the relief capability of the remaining General Electric BWR/6 STS B 3.4.4-3 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.4 S/RVs, the low probability of an event requiring S/RV actuation, and a reasonable time to complete the Required Action. [Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.]

B.1

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REVIEWERS NOTE ----------------------------------

Adoption of a MODE 3 end state requires the licensee to make the following commitments:

1. [LICENSEE] will follow the guidance established in Section 11 of NUMARC 93-01, "Industry Guidance for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Nuclear Management and Resource Council, Revision [4F].

2. [LICENSEE] will follow the guidance established in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 2, "Technical Specifications End States, NEDC-32988-A," November 2009.

If the inoperable required S/RV cannot be restored to OPERABLE status within the associated Completion Time of Required Action A.1, the plant must be brought to a MODE in which overall plant risk is minimized. To BASES ACTIONS (continued) achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 4) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.

Required Action B.1 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 3. This Note prohibits the use of LCO 3.0.4.a to enter MODE 3 during startup with the LCO not met.

However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 3, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

General Electric BWR/6 STS B 3.4.4-4 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.4 The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

AC.1 and AC.2 If the OPS is inoperable, [two] or more [required] S/RVs are inoperable, a transient may result in the violation of the ASME Code limit on reactor pressure. The plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This Surveillance verifies that the OPS has the capability to prevent the reactor steam dome pressure from exceeding Safety Limit 2.1.2.

demonstrates that the [required] S/RVs will open at the pressures assumed in the safety analysis of Reference 2. The testing of the demonstration of the S/RV safety mode function lift settings is must performed during shutdown, since this is a bench test[, to be done and in accordance with the INSERVICE TESTING PROGRAM]. The measured S/RV mechanical lift pressures determined in accordance with the INSERVICE TESTING PROGRAM are reviewed and compared to the overpressure analysis to verify that the collective performance of the credited S/RVs will ensure Safety Limit 2.1.2 is protected.

Should one or more of the credited S/RVs not actuate within the assumed tolerance during as-found testing, the actual lift values will be used to evaluate the affected previous cycle overpressure analyses to determine whether the Safety Limit was protected. In this case, the SR is met by a combination of testing and calculation.

The effect on OPS operability of S/RVs not actuating within the assumed tolerance during as-found testing will be evaluated for the current cycle under the corrective action program using the NRC-approved methods for overpressure and accident analyses.

The lift setting pressure shall BASES SURVEILLANCE REQUIREMENTS (continued) correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is +/- [3]% for OPERABILITY; however, the valves are reset to +/- 1% during the Surveillance to allow for drift. [A Note is provided to allow up to [two] of the required [11] S/RVs to be physically replaced with S/RVs with lower General Electric BWR/6 STS B 3.4.4-5 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.4 setpoints. This provides operational flexibility which maintains the assumptions in the over-pressure analysis.]

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REVIEWERS NOTE-----------------------------------

If the testing is within the scope of the licensee's INSERVICE TESTING PROGRAM, the Frequency "In accordance with the INSERVICE TESTING PROGRAM" should be used. Otherwise, the periodic Frequency of 18 months or the reference to the Surveillance Frequency Control Program should be used.

[ The [18 month] Frequency was selected because this Surveillance must be performed during shutdown conditions and is based on the time between refuelings.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

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]

SR 3.4.4.2 The OPS assumes that tThe [required] relief function mode S/RVs are required to actuate automatically upon receipt of specific initiation signals.

A system functional test is performed to verify the mechanical portions of the automatic relief function mode operate as designed when initiated either by an actual or simulated initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.5.4 overlaps this SR to provide complete testing of the relief mode safety function.

BASES SURVEILLANCE REQUIREMENTS (continued)

[ The [18 month] Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the SR when performed at the [18 month]

Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

General Electric BWR/6 STS B 3.4.4-6 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.4 OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

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]

This SR is modified by a Note that excludes valve actuation. The SR may be performed by removing the actuator and verifying its operation. This prevents an RPV pressure blowdown.

SR 3.4.4.3 A manual actuation of each [required] S/RV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the S/RVs divert steam flow upon opening.

Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at which this test is to be performed is 950 psig (the pressure recommended by the valve manufacturer). Adequate steam flow is represented by [at least 1.25 turbine bypass valves open, or total steam flow 106 lb/hr]. Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required General Electric BWR/6 STS B 3.4.4-7 Rev. 5.0

TSTF-576, Rev. 1 OPSS/RVs B 3.4.4 BASES SURVEILLANCE REQUIREMENTS (continued) pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. If the valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the S/RV is considered OPERABLE.

[ The [18] month on a STAGGERED TEST BASIS Frequency ensures that each solenoid for each S/RV is alternately tested. The 18 month Frequency was developed based on the S/RV tests required by the ASME (Ref. 1). Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

OR The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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REVIEWERS NOTE-----------------------------------

Plants controlling Surveillance Frequencies under a Surveillance Frequency Control Program should utilize the appropriate Frequency description, given above, and the appropriate choice of Frequency in the Surveillance Requirement.

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]

REFERENCES 1. FSAR, Section [5.2.5.5.3].

21. ASME Code for Operation and Maintenance of Nuclear Power Plants.

2. FSAR, Section [5.2.5.5.3].

3. FSAR, Section [15].

4. NEDC-32988-A, Revision 2, Technical Justification to Support Risk-Informed Modification to Selected Required End States for BWR Plants, December 2002.

General Electric BWR/6 STS B 3.4.4-8 Rev. 5.0

TSTF-576, Rev. 1 Reactor Steam Dome Pressure B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.12 Reactor Steam Dome Pressure BASES BACKGROUND The reactor steam dome pressure is an assumed initial condition of Design Basis Accidents (DBAs) and transients and is also an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria.

APPLICABLE The reactor steam dome pressure of [1045] psig is an initial condition of SAFETY the vessel overpressure protection analysis of Reference 1. This ANALYSES analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the Overpressure Protection Systempressure relief system, primarily the safety/relief valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved.

Reference 2 also assumes an initial reactor steam dome pressure for the analysis of DBAs and transients used to determine the limits for fuel cladding integrity MCPR (see Bases for LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)").

Reactor steam dome pressure satisfies the requirements of Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The specified reactor steam dome pressure limit of [1045] psig ensures the plant is operated within the assumptions of the transient analyses.

Operation above the limit may result in a transient response more severe than analyzed.

APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these MODES, the reactor may be generating significant steam, and the DBAs and transients are bounding.

In MODES 3, 4, and 5, the limit is not applicable because the reactor is shut down. In these MODES, the reactor pressure is well below the required limit, and no anticipated events will challenge the overpressure limits.

General Electric BWR/6 STS B 3.4.12-1 Rev. 5.0

TSTF-576, Rev. 1 LLS Valves B 3.6.1.6 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.6 Low-Low Set (LLS) Valves BASES BACKGROUND The safety/relief valves (S/RVs) can actuate either in the relief mode or, the safety mode as part of the Overpressure Protection System, the Automatic Depressurization System mode, or the LLS mode. In the LLS mode (or power actuated mode of operation), a pneumatic diaphragm and stem assembly overcome the spring force and open the pilot valve.

As in the safety mode, opening the pilot valve allows a differential pressure to develop across the main valve piston and thus opens the main valve. The main valve can stay open with valve inlet steam pressure as low as [0] psig. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure exceeds the safety mode pressure setpoints.

[Six] of the S/RVs are equipped to provide the LLS function. The LLS logic causes the LLS valves to be opened at a lower pressure than the relief or safety mode pressure setpoints and stay open longer, such that reopening of more than one S/RV is prevented on subsequent actuations.

Therefore, the LLS function prevents excessive short duration S/RV cycles with valve actuation at the relief setpoint.

Each S/RV discharges steam through a discharge line and quencher to a location near the bottom of the suppression pool, which causes a load on the suppression pool wall. Actuation at lower reactor pressure results in a lower load.

APPLICABLE The LLS relief mode functions to ensure that the containment design SAFETY basis of one S/RV operating on "subsequent actuations" is met (Ref. 1).

ANALYSES In other words, multiple simultaneous openings of S/RVs (following the initial opening) and the corresponding higher loads, are avoided. The safety analysis demonstrates that the LLS functions to avoid the induced thrust loads on the S/RV discharge line resulting from "subsequent actuations" of the S/RV during Design Basis Accidents (DBAs).

Furthermore, the LLS function justifies the primary containment analysis assumption that multiple simultaneous S/RV openings occur only on the initial actuation for DBAs. Even though [six] LLS S/RVs are specified, all

[six] LLS S/RVs do not operate in any DBA analysis.

LLS valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO [Six LLS valves are required to be OPERABLE to satisfy the assumptions of the safety analysis (Ref. 2). The requirements of this LCO are applicable to the mechanical and electrical/pneumatic capability of the LLS valves to function for controlling the opening and closing of the S/RVs General Electric BWR/6 STS B 3.6.1.6-1 Rev. 5.0